ML20214W899

Application for Renewal of License SNM-1405

ML20214W899
Person / Time
Site:	07001359
Issue date:	05/20/1987
From:	IRT CORP.
To:
Shared Package
ML20214W892	List: ML20214W889 ML20214W899
References
IRT-4171-012, IRT-4171-12, NUDOCS 8706160237
Download: ML20214W899 (127)

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IRT 4171-012 APPLICATION FOR RENEWAL OF SPECIAL NUCLEAR MATERIAL LICENSE SNM-1405 Submitted to g wgqgg UNITED STATES l 7 ' + y i 3 l NUCLEAR REGULATORY COMMISSION !? s .k IRT CORPORATION Lj.',._i ] E '- San Diego, California -e# IRTCORPORATION April 20,1987 3030 Caltan Road. P O Box 85317 ( San Diego Cabfornia 92138 619/450 4343. Telex. 69-5412 -)

3 s CONTENTS 4 4 1. IDENTIFICAflON OF APPLICANT................................... I

2. -

LOCATIONS WHERE SPECIAL NUCLEAR MATERIAL WILL BE USED 3 3. POSSESSION LIM ITS................................................ 5 3.1 Total License Li mit............................................ 5 3.2 Form of Licensed Material...................................... 6 3.3 Limitations and Exemptions..................................... 7 3.3.1 Specific Limitations for Areas Not Equipped With Criticality Monitors 7 3.3.2 Specific Limitations for the 8221 ArJons R oad Facili t ie s......................................... 8 i 3.3.3 Exemption from Requirements Set Forth in 10 C F R 7 0.2 4.......................................... 8A 3.3.4 Exemption for the Arjons Warehouse Facility............... 8B 9 4. PROPOSED USES 4.1 Development of Nondestructive Inspection and Assay Equipment..... 9 4.1.1 Nuclear Fuel Quality Control Equipment................... 9 4.1.2 Waste Management Equipment 10 4.1.3 Safeguards Equipment................................... 10 l 4.1.4 Research and Development Programs 10 5. R ADIATION 5AFETY ORG ANIZ ATION................................ 11 j 6. ADMINISTRATIVE PROCEDURES.................................... 27 ? i 6.1 Project Authorization Procedures................................ 27 4 6.1.1 Radiation Work Authorization Request Procedures.......... 27 6.1.2 Radiation Safety Of ficer Review Procedures............... 28 6.1.3 Radiation Safety Committee Review Procedures............ 30 i 6.2 Project Changes and Renewals.................................. 31 i 6.3 Selection Criteria for Health Physics Staff and Radiation l Safety Com mit tee............................................. 32 6.4 Operations of the Radiation Safety Officer and Health Physicist..... 32 6.3 Internal Inspection and Review.................................. 33 6.6 Operations of the Radiation Safety Committee.................... 34 6.7 Control and Accountability of Special Nuclear Material............. 34 6.7.1 Organization for Control, Safeguarding and Accountability of 5NM 34 l April 20,1987 { Pageill 3

r 0 6.8 Storage of Special Nuclear Materials............................. 35 6.8.1 3030 Callan Road Storage Vault 35 6.8.2 S to rage Crite ria........................................ 35 6.8.3 -Arjons Warehouse Storage 38 6.8.4 Temporary Storage of SNM in Use 39 6.8.5 General Procedures for Use of Special Nuclear Material 39 6.9 Progra m R ecords.............................................. 43 7. RADIATION PROTECTION PROCEDURES 45 7.1 Radiation Protection Manual.................................... 45 46 7.2 Personnel Monitoring 7.2.1 Film Badges 46 7.2.2 Bloassay 47 7.2.3 Whole-Body Monitoring.................................. 49 7.3 Limits of Radiation in Controlled and Uncontrolled Areas........... 50 7.3.1 Co n trolled A rea........................................ 50 7.3.2 Uncontrolled Areas..................................... 50 7.3.3 Contamination Control.................................. 51 7.3.4 Radiological Survey..................................... 52 7.3.5 Of f-Site Opera t ions..................................... 53 8. INSTR UCTION O F PE RSON NE L...................................... 55 8.1 Formal Training for New Users.................................. 55 8.2 Pe r iodic R e t r aining............................................ 55 9. TECHNICAL CA PABILITIES........................................ 57 9.1 General Purposes o f Use........................................ 57 9.2 Organizational Structure 57 9.3 Technical Personnel 59 9.4 Fa c i l i t i e s..................................................... 71 9.4.1 3 0 3 0 Callan Roa d....................................... 71 9.4.2 8221 Arjons Road 76 9.5 Effluent Control 76 9.5.1 Air................................................... 76 9.5.2 Liquid 76 9.6 In st ru m en ta t ion............................................... 78 9.6.1 Personnel Monitoring Devices............................ 78 9.6.2 Radiation Monitoring and Survey Instruments............... 78 9.6.3 Radioactive Material Assay.............................. 78 81 9.6.4 Air Samples April 20,1987 Pageiv

nU 10. MANUFACTURING AND QUALITY ASSURANCE PROCEDURES......... 83 11. W AST E D IS PO S A L.................................................. 85 1 1.1 Solid Waste Disposal........................................... 85 !!.l.1 Work Area Waste Receptacles............................ 85 11.1.2 Collection of Wa ste..................................... 85 11.1.3 Monitoring ............................................ 86 11.1.4 Waste Collection and Storage Area 86 11.1.5 Packaging of Shipping Drum s............................. 86 !!.l.6 Monitoring Shipping Drums .............................. 86 11.1.7 Disposal o f Solid Was t e.................................. 87 11.2 Liquid Wa ste Disposal.......................................... 87 11.2.1 Work Area Liquid Waste Receptacles...................... 87 !!.2.2 Monitoring and Collection of Liquid Waste................. 87 12. C E R TI T I C A T E..................................................... 89 APPENDIX I: SUPPLEMENTAL INFORMATION CONCERNING PROPOSED USES .................................... 91 APPENDIX !!: R ADIATION WORK AUTHORIZATION FORM............ 95 V APPENDIX 111: WORKPLACES FOR UNSEALED RADIONUCLIDES....... 99 APPENDIX IV: OUTLINE - IRT R ADIOLOGICAL SAFETY COURSE IRT RADIOLOGICAL MANUAL IRT 4141-009........... 109 APPENDIX V: SUPPLEMENTAL INFORMATION FOR THE USE OF UNSEALED SOLID MATERIAL IN NONDISPERSIBLE FORM..............................................I15 1 1 O \\ \\ G April 20,1987 Page v

O Ws page Intentionally lef t blank. O l i l l t l 9 April 20,1987 Pagev1 ~~"'""~vmm

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1. IDENTIFICATION OF APPLICANT This license renewal application is made by IRT Corporation and authorizes activities which require licensing of operations involving the use of radioactive materials.

( The headquarters of IRT Corporation are located in San Diego, California at the following address: 1. Name of Applicants IRT Corporation 2. Malling Address m P.O. Box 85317 San Diego, California 92138-5317 3. Telephone Number, Radiation Safety Office: Area Code: 619 l Number: 450-4343 l l IRT Corporation, formerly Intelcom Rad Tech, came into being in April 1973. Prior to that date, all of the operating components and facilities of IRT comprised the Radiation Technology Division of General Atomic Company in San Diego, California. Prior to 1967, the components of General Atomic comprised the General Atomic Division of General Dynamics Corporation. Within various internal organizational structures, the components of Rad Tech had been licensed and operating effectively and safely for nearly 25 years. For the past 14 years, since the Company's separation from General Atomic, IRT has operated under State of California Radioactive Materials Licenses 2468-59 and 2468-80, and U. 5. Nuclear Regulatory License SNM 1405. OV April 30,1987 Page1

IRT Corporation is financially quallfled to conduct the operations requested to be licensed and is financially quallfled to be responsible for any cleanup and decontamina-tion required if any of its licensed operations are terminated and facilities opened for general use. The applicant, IRT Corporation, is a corporation which was incorporated in the state of Delaware. The applicant has no known control or ownership by any allen, foreign corporation, or foreign government. O O April 30,1987 Page 2

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2. LOCATIONS WHERE SPECIAL NUCLEAR MATERIAL WILL BE USED This license application requests authorization to use special nuclear materials at the following sites:

1. Laboratories and facilities at the IRT headquarters located at 3030 Callan Road, San Diego, Californla 92121. 2. Manufacturing facilities located at 8221 ArJons Road, San Diego, California 92126. 3. Temporary job sites of the !!censee anywhere in the United States where the Nuclear Regulatory Commission maintains jurisdiction. I O April 30,1987 Page 3

O This page Intentionally lef t blank. O I i l April 30,1987 Page 4

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3. POSSESSION LIMITS 3.1 TOTAL LICENSE LIMIT 6.

Material (Enrichment)

7. Form

8. Limit (grams)

A. Plutonium-239 (75%) Scaled sources 30 B. Plutonium-239 (75%) Mixed oxide fuel rods 365 as sealed sources C. Plutonium-239 (75%) Any 1 D. Plutonium-239 (75%) PuBe Neutron Sources 33 j E. Plutonium-238 (80%) PuBe Neutron Sources 2 F. Uranium-235 (>20%) Scaled Sources 200 j q G. Uranium-235 (>20%) Any 5 kJ H. Uranium-235 (10%-20%) Scaled Sources 200 Uranium-235 (10%-20%) Any 5 3. Uranium-235 (<10%) Mixed oxide fuel rods 284 as sealed sources K. Uranium-235 (<10%) Scaled Sources 1400 L. Uranium-235 (<10%) Non-Dispersable solid 300* M. Uranium-235 (<10%) Any 10

• This 300 grams of non-dispersable solid material is included as a part of the 1400 gram limit for the <10% enr. U-235.

The quantity of unsealed U-235 and Pu and any mixture thereof shall be less than the values set forth in 10 CFR 150.11. l Excluding the'Pu/U mixed oxide fuel rods that are possessed for storage only, the total quantity of SNM at any one time shall not exceed the amount specified in 10 CFR 73.2 (y), "Special Nuclear Material of Low Strategic Significance." I l OV April 30,1987 Page5 l

,V ,i i s, l s 9a. Authorized Use Items A, C, D, E, F, G, H, I, 3, K, and M are to be used for the development and tes' ting of nondestructive detection, assay, and inspection equipment; for development i; and. testing of nuclear fuel, waste management, and safeguards equipment; and for research and development; in accordance with the statements, representations, and , tc'onditions sp'ecified in this license application. E s Iterns B and 3 are to be used for storage only. I 9b. Authorized Places of Use s Items A, C, D, E, F, G, H, I, 3, K, and M will be used at 3030 Callan Road San E Diego, Califo nia 92121. Items F, H, and K (within,the limitations specified in paragraph 3.3.2) will be used E -at the Manufacturing Facility located at 8221 Arjons Road, Suite F San Diego, California '92126. Itemit and 3 will be stored at the facility located at 8221 Arjons Road, Suite F I s San Diego,'Califo~rnia 92L26. 3.2 FORM OF LICENSED MATERIAL The SNM specified in the possession limits takes a variety of physical and chemical forms.' These include, bt.t are not limited to: Metal plates and rods Metal foils Alloys Oxide and carbide pellets Oxide Powders Graphite-coated particles Mixed oxide fuel rods Uranium / aluminum alloy fuel plates Assorted contal ers with small quantitles of SNM ( Liquids containing microcurie quantities of SNM. The majority of the SNM will be in the form of sealed sources, i.e., fuel rods, fuel plates and sealyd metal or plastic containers. O April 36,1987 Page 6 l s i

.y-() 3.3 LIMITATIONS AND EXEMPTIONS 3.3.1 Specific Limitations for Areas Not Equipped With Criticality Monitors The following limitations shall be established as the maximum inventory of SNM a!! owed in each facility or area that is not equipped with criticality monitors. All transfers of SNM into or out of each area shall be under the control of the Radiation Safety (Accountability) Office. These limitations shall also be the inventory limits for the 3030 Callan Road Facility. Material (Enrichment) Form Limit (grams) Plutonium-239 (75%) Sealed Sources 30 Plutonium-239 (75%) Any 1 Plutonium-239 (75%) PuBe Neutron Sources 33 Plutonium-238 (80%) PuBe Neutron Sources 2 E Uranium-235 (>20%) Sealed Sources 200 -O Q Uranium-235 (>20%) Any 5 Uranium-235 (10 to 20%) Sealed Sources 200 Uranium-235 (10 to 20%) Any 5 Uranium-235 (<10%) Sealed Sources 1400 Uranium-235 (<10%) Non-Dispersable solid 300* Uranium-235 (<10%) Any 10

• This 300 grams of nondispersable solid material is included as a part of the 1400 gram limit for the <10% enr. U-235.

The quantity of 'U is restricted to an amount not to exceed that defined by the following formula: Mass U-235 (>10%) Mass U-235 (<10%) 41 400 1400 k April 30,1987 Page 7

The limits described above will apply to each of the experimental areas provided that they are separate rooms and are isolated from adjoining use areas by one of the following criteria: 1. Surface-to-surface separation of at least 12 feet. 2. Separation by at least the nuclear equivalent of eight inches of concrete of a density of 140 pounds per cubic foot. This isolation criteria was obtained from TID-7016, Revision 1, pages 26 and 27, and applies to those systems as described in Table IV with the stipulation that more reactive systems would require 12 inches of concrete. Considering the specified mass limits and their most reactive configuration as the minima in the critical mass curves in Figures 8, 34, and 27 of TID-7028 (H. C. Paxton, et al., " Critical Dimensions of Systems 235 233 "), the mass safety factors for 235U,239Pu and U 233 Containing U,239Pu and U are 1.6, 2.0 and 1.92, respectively. Those systems specified in Table IV (TID-7016) have mass safety factors of the order of 1.33,1.43 and 1.55, respectively. The specified mass limits for these areas clearly fall within the limits described in Table IV and the isolation criteria is therefore applicable. 3.3.2 Specific Limitations for the 8221 Arjons Road Facility A. Pu/U Test Rod Storage Vault E The material in this facility is for STORAGE ONLY. The material that will be stored is coatained in 13 mixed oxide Pu/U fuel rods. Material (Enrichment) Form Limit (grams) Plutonium-239 (75%) Sealed Sources 365 Mixed oxide fuel rods Uranium-235 (<10%) Sealed Sources 284 Mixed oxide fuel rods i l April 30,1987 Page3 l

, pC B. Manufacturing Facility Material (Enrichment) Form Limit (grams) Uranium-235 ( >20%) Sealed Sources <l5 Uranium-235 (10 to 20%) Scaled Sources <l5 Uranium-235 (<10%) Scaled Sources <15 The total amount of Uranium will not exceed the following formula: Mass U-235 (>20%) Mass U-235 (10 to 20%) Mass U-235 (<10%) + +

1. I*

15 15 15 3.3.3 Exemption from Requirements Set Forth in 10 CFR 70.24 S The mass limitation given in Section 3.3.1 for U when coupled to the maximum 233 weighted mass of Pu and 0 (weighting factor of 2.5) is below critical mass for homogenized systems of various enrichments with optimum moderation and full reflection as given in Figure 13 of TID-7028. The materials are not in a form to be homogenized and there are no massive moderators or reflectors of beryllium, heavy water or graphite present in the storage or work areas. IRT, therefore, requests exemption from the requirements set forth in 10 CFR 70.24 for all areas not exceeding these mass limitations. (D ~ Q April 20,1987 Page 8A i

3.3.4 Exemption for the Arjons Storage Vault Facility I The total amo_unt of materialinvolved is: 365 grams of Plutonium, 277 grams of 2.1% enriched Uranium-235, and 1174 grams of depleted Uranium containing 7 grams of U-235. This materialis contained in 13 sealed stainless steel fuel rods. The rods will be stored I in two boxes that will be placed end-to-end inside the storage container. Box No. I will contain six rods with: 161.5 gm of Pu and 138.5 gm of U-235 (2.1% enr.). Box No. 2 will contain seven rods with: 203.5 gm Pu, 138.5 gm U-235 (2.1% enr.), and 7.0 gm U-235 contained in 1174 gm dep-U. The internal measurements of the boxes are 10.5 cm x 10.5 cm x 195 cm long. The total internal volume of each box is 21,500 cubic centimeters or 21.5 liters. Box No. 2, which has the most Plutonium, will have a Pu concentration of 9.5 gm/ liter. If all of the SNM were plutonium, the concentration would be 16.2 gm/ liter. Referring to TID-7028 Figure 19, " Estimated Critical Diameters of Infinite Cylinders of Homogen-eous Water-Moderated Plutonium," the graph shows that a water reflected metal-water mixture in a 4-inch (10 cm) diameter cylinder would require a concentration of 3.2 Kg/ liter to achieve a critical mass. From this we calculated that the Plutonium in storage will have a safety factor of 336, and the " total SNM" will have a safety factor of 197. The materialis not in form to be homogenized. The boxes will be stored inside a steel container sealed with bolts and an "O" ring. There are no massive moderators or reflectors in or around the storage cask. IRT, therefore, requests exemption from the requirements set forth in 10 CFR 70.24 for the Arjons Warehouse SNM Storage Facility. April 30,1987 Page 8B

s

4. PROPOSED USES The SNM authorized by this license will be used for development of nondestruc-tive inspection and assay equipment, and for Research and Development as defined in 10 CFR 70.4(j).

4.1 DEVELOPMENT OF NONDESTRUCTIVE INSPECTION AND ASSAY EQUIPMENT The nondestructive testing and assay equipment includes instruments for nuclear fuel quality control, waste management and safeguards. 4.1.1 Nuclear Fuel Quality Control Equipment Some typical instruments developed in these programs are described below. OV 1. The Active Fuel Rod Scanner (AFRS-110) is a quality control tool for measuring the enrichment uniformity, total fissile content, rod and plenum length, fuel stock length, and other important fuel rod parameters. The instrument tests the rods by low-intensity irradiation with neutrons from a 252Cf source followed by delayed gamma-ray detection of the resulting fissions. 2. The Magnetic Induction Rod Scanner (MIRS) is used to determine the Cd (burnable poison) content of individual fuel pellets inside a sealed fuel rod. u 3. The Isotopic Source Assay System (ISAS) is used for assay of fissile or fertile materials within containers up to 8 inches in diameter and 51 inches in 252 length. This instrument uses Cf neutrons to irradiate the sample and automatically counts the resultant gamma rays. 4. The Isotopic Source Adjustable Fissometer Assay System (ISAF) uses a Am Li neutron murce to irradiate the sample and determine the fissile content of materials in containers up to five gallons in volume. C'i V April 20,1987 Page 9 --a, -,-m- --n.., ,,,,e. - - - -

4.1.2 Waste Management Equipment Some typicaLwaste management equipment used in this program is described below. 1. The Fission Multiplicity Detector (FMD) was developed by IRT for the passive assay of plutonium content in low-level hydrogenous waste materials con-tained in 55-gallon drums. 2. The High-Level Neutron Coincidence Counter (HLNCC-100) is a portable neutron coincidence detection system for assaying plutonium compounds. 3. The Dual Range Neutron Well Coinciden' e Counter (DRC-100) was developed c to provide accurate determination of plutonium for a wide range of masses. 4.1.3 Safeguards Equipment The primary safeguard device developed at IRT is the PRM-100 Portal Radiation Monitor. This device detects both special nuclear materials and byproduct materials moving through the portal at normal walking speed. O 4.1.4 Research and Development Programs 1. Experimental measurements involving nuclear parameters such as delayed gamma rays and fission yield. 2. Calibration of instrumentation including alpha, beta, gamma-ray, neutron and fission fragment detectors. The SNM used in these two programs includes 235 sealed or encapsulated Pu and U of generally less than 200 grams, and milligram quantities of Pu in the form of foils. '). Radiochemical analysis using submicrocurie quantities of Pu and 'O solutions. Appendix I contains general procedures for safety and handling of unsealed SNM for research and development programs. O April 20,1987 Page 10

'( V

5. RADIATION SAFETY ORGANIZATION The IRT radiation safety organization consists of a Radiation Safety Officer (RSO) who also serves as the Criticality Safety Officer (CSO), the Radiation and Criticality Safety Committee (RSC), and the Health Physics Staff.

The RSC is composed of not less than five qualified members, including the RSO. The committee represents both management and operating groups. The duties and functions of this committee are taken from American National Standard " Administrative Practices for . Nuclear Criticality Safety," ANSI /ANS-8.19-1984. The purpose of the radiation safety organization is to assure compliance with license requirements regarding the use of licensed materials and equipment; to assure fm compliance with established radiological safety standards and the nuclear Criticality b, Safety Program; to administer SNM safeguards and accountability; and to provide u license administration. This organization establishes policies necessary for safety of radiation operations and publishes them in guides and manuals; provides various services such as personnel monitoring, dose rate measurement, radioactive material detection and assay, air and water sampling, environmental monitoring, and instructional and training programs. It provides interpretation of licenses, preparation and processing of license applications, dissemination of license requirements, and maintenance of master license records. The RSC acts in both a review function and an audit function. It is responsible for the critical review of all radiation-related work within the Company and must give authorization for all such activities. The Committee also audits all work involving licensed materials and radiation-producing machines for conformance to and effective-ness of all applicable procedures and practices and regulations. These functions are carried out through the review of experimental plans and equipment, and are conducted to assure that license conditions are satisfied and that all reasonable precautions are taken to avoid accidental criticality in the handling and storage of SNM. g It is the policy of the radiation safety organization to assure that all approved V operations are conducted in such a manner so as to keep personnel exposures, radiation April 20,1987 Page 11

levels, and the release of airborne and liquid effluents to a level that is as low as is reasonably achievable. The goal is to maintain personnel exposures and the release of effluents to a value less than 25 percent of those specified in 10 CFR 20. Where monitoring programs indicate that these levels are routinely being exceeded, the operation will be reviewed to determine if it is practical to reduce the levels. At no time will the limits specified in 10 CFR 20 be exceeded. Committee members are selected in accordance with Section 6.3. At least two members are selected for their experience and familiarity with nuclear criticality. A listing of the members of the RSC and statements of their training and experience, Form RH 2050 A, follows. O O April 20,1987 Page 12

O Radiation Safety Officer and Chief Principal P. R. Maschka Health Physicist S. 3. Friesenbahn - Physicist Radiation Safety Title Radiation Safety Title Officer - Committee, Chairman . Principal Test K. L. Crosbie Engineer

3. W. Harrity Group Leader Alternate Radiation Title Radiation Safety Title Safety Officer.

Committee, Member ' Radiation Safety Committee Test Group Linac Facility D. E. Willis Leader

3. C. Young Manager Radiation Safety Title Radiation Safety Title Committee, Member Committee, Member Member Title Member Title

\\ September 21,1983 Page 13 1 . - -... -, - -,,. ~, - -,. - - - - -. - - - - - - - - - - - - - - - - - -. - - - - - - -

O State of CalifoEs Radiologie Hemith Seenon 744 P Street Department of Health Sacramento. CaWorne 96814 STATEMENT OF TRAINING AND EXPERIENCE (Use additional sheets as necessr.ry) Instruction: Every individual proposing to use radioactive material is required to submit a Statement of Training and Experience in duplicate to the address given above. Physicians should request Form RH 2000 when applying for human use authorizanons. Radiation Safety Officer and Chief 1. Name of proposed user: Paul R. Maschka Position title: Health Physicist Address': 2111 Weatherby Avenue City: Escondido Zip: 92027 To be included on Lic. No. 2468-80 in name of IRT Corocration 2. Description of proposed use Research and development; neutron radiography, gauging and assaying systems, instruc-tion and demonstraticn 3. Training: a. High School Graduate: Yes X No b. Colkge or University: Name and location Creighton University. Omaha. Nebraska Years completed 2 Degree none Course of study c. Education specifically applicable to use of radioactive material U. S. Army Nuclear Power Plant Training Course (48 weeks) Ft. Belvoir, Virginia General Atomic Radiological Safety Course - passed 4. Experience: a. List experience with radioacuvity beginning with most recent (1) Dates: From M3Y 1973 to present Title and duties: Health Physicist, responsible for administration and all operational radiation safety for IRT Employer. IRT Corporation Address: P.O. Box 80817. San Diego, CA 92138 (2) Dates: From Dec. 1963 to April 1973 Title and duties: Health ohysics surveyor, monitored radio-chem lab. hot cell, reactor facility, and linear accelerator Employer: General Atomic Address: San Ofego, CA (3) Dates: From Jan. 1960 to Dec. 1963 Title and duties: Health ohysics tech. and reactor operator (SM.1) at Nuclear Power Plant, Ft. Belvoir, Va., & PM-3A Nuclear Power Plant at McMurdo, Antartica Employer: U. S. Air Force Address: Ft. Belvoir, VA RH M4 (7/76) O April 20,1987 Page 14

rf ( Y' w. b. Radioactive materials previously used. Cite typical radioisotopes in appropnate box and key to Part 4.a above: Quantities Handled Microcurg ,Wilicuries ,, Curic.s., Kilocuries Co*3$ Cs "239 Gs"' E 241 Ca50, Sr90 Go " Co". Gs -'. Scaled sources 2 10 U Pu cd g ef,pgp 235 233 235 239 235 232 Unsealed alpha g 241,U 23}Pu 238. Th 3 U 237 enutters 238 g Np Th ,g U Unsealed beta-Sr90 Cs137 MFP & MAP MFP MFP gamma emitters h22. Ba133 109 Cd gp PuSe, PcSe Neutron sources pol 1, Cf-252 Cf-252 Cf-252 Describe procedures similar to those proposed in Part 2 with which you have had experience. c. Indicate months or years for each and key to Part 4.a above. Research and development 4a (1) - May 1973 to present 4a (2) - Dec 1963 to April 1973 .d d. Indicate which types of facilities you have used and key to Part 4.a. (X) Ordinary Chemical laboratories 44 1,2,3 (X) " Controlled Area" (Type B) laboratories 4a 2,3 (X) Glove boxes 4a 1,2 (X) Shielded giove boxes 4a 2 (X) Caves with remote menipulators 4a 2 (X) Field operations with portable equipment 4a 1,2,3 5. Certificate: I hereby artify that all information contained in this Statement is true and correct. A / J W A l'la,. W L,2./ d / 9 ?~/ Signature of proposse user Oste April 20,1987 Page 15

O State of Caliform Radiologic Me:Hh Section 744 P Street Depertment of Health 8**"*"# N'"" STATEMENT OF TRAINING AND EXPERIENCE (Use additional sheets as necessary) Instruction: Every individual proposing to use radioacuve material is required to submit a Staternent of Training and Experience in duplicate to the address given above. Physicians should request Form RH 2000 when applying for human use authorizauons. 1. Name of proposed user: Stanley J. Friesenhan Position title: Physicist Address: 12906 Conley City: Poway Zip: 92064 To be included en Lic. No. 2468-PO in name of IRT Corporation 2. Description of proposed use Research and development 3. Training: a. High School Graduat : Yes X No b. College or Universityi Name and location University of Notre Dame. South Bend. Indiana Years completed 6 Degree MS Course of study Physics c. Education specifically applicable to use of radioactive material 4 Experience: Ilst experience with radioactivity beginning with most recent s. (1) Dates: From 1961 to cresent Title and duties: Principal ohysicist. research and development Employer. [RT Corporation Address: P.O. Box 80817. San Dieco. CA 92138 (2) Dates: From 1957 to 1960 l Title and duties: Physicist, research and development Employer: Hanford labs Address: Richland, Washington (3) Dates: From to l Title and duties: i 8mployer: Address: aw tema one O April 20,1987 Page 16 l

j-%., (J b. Radsoactive rnaterials previously used. Cite typical radioisotopes in appropnate box and key to i Part 4.a above: Quantities Handled Microcuries Millicunes Curies Kilocunes 4a 1,2 4a 1,2 Scaled sources 13 7,. _ 60r_ Unsealed alpha 4a 1,2 ww 137,_ 60,.. v,g v*5 4a 1,2 4a 1 emitters 239po,235g py, U 239g,235,. 239 234 Unsealed beta-4a 1 4a 1 gamma enitters MAP & MFP MAP & MFP 4a 1,z 4a 1 4a 1.4 Neutron sources Cf-252 Cf-252 Cf-252. PuBe Describe proadures similar to those proposed in Part 2 with which you have had experience. c. Indicate months or years for each ar.d key to Part 4.a above. Research and development 4a 1,2 - 24 years (D^ t,j d. Indicate which types of facilities you have used and key to Part 4.a. (4 Ordinary Chemical laboratories 4a 1,2 ($ " Controlled Ares".(Type B) laboratories 4a 1,2 (4. Clove boxes da 1,2 - () Shielded glove boxes () Caves with remote manipulators (4 Field operations with portable equipment 4a 1,2 5. Certificate: I hereby crstify that all information antained in this Statement is true and correct. 2/12 A, 47. L L ? - 7 V -d'/ y / sisa.tur. or propo a u e om O v April 20,1987 Page 17

O1 State of Californtr Radiologic Hastth Secnon 744 P street Department of Hegith Sacramento. Cairfornia 95814 l STATEMENT OF TRAINING AND EXPERIENCE (Use additional sheets as necessary) Instruction: Every individual proposing to use radioactive material is required to submit a Statement of Training and Expenence in duplicate to the address given above. Physicians should request Form RH 2000 when applying for human use authorizzuons. Alternate Radiation 1. Name of proposed user: Kay L. Crosbie Position title: Safety Officer Address: 5002 Northaven Avenue City: San Diego Zip: 92110 To be included on Lic. No. 2468-80 in name of f RT Corcoration i 2. Description of proposed use a. Research and Development as defined in CRCR Section 30175(j) b. Nondestructive testing, radiagraphy, gauging and assay systems, and c. Instruction and demonstration. 3. Training High School Graduate: Yes XX No Pratt Inst., Brooklyn, N.Y.; a. b. College or University: Name and location Univ. of VA, Charlottesville, VA Years completed 6 Degree MNE Course of study Chemical Engrng; Nuclear Engrng. c. Education specifically applicable to use of radioactive material Experimental Nuclear Engineering Lab. Radiation Shielding and biological effects of radiation at University of Virginia 4 Experience: a. IJst experience with radioactivity beginning with most recent (1) Dates: From April 1973 to present Title and duties: Radiation safety officer, principal physicist--resconsible for radiatic safety nealtn pnysics, licensing and bNi account 10141ty; K60 witn suDcritical as w b, lies and NDT Employer. IRT Corporation Address. P.O. Box 80817, San Diego, CA 92138 (2) Dates: From Sept. 1963 to April 1973 Title and duties: Staff Engineer--experimental research; supervisor of accelerator pulsed reactor; R&D with linear accelerator and nondestructive testing Employer: Gulf Rad Tech Address: p.0. Box 81608, San Diego, CA (3) Dates: From Sept. 1961 to Sept. 1963 Title and duties: Staff member; pulsed reactor supervisor--responsible for reactor ~ I operations and safety Employer: Sandia Corcoratinn Address: Albu %.rm!. New wtxico MM 20g4A 17176) 9 April 20,1987 PageIS

'f %W W i D' b. Radioacuve materials previously used. Cite typical radioisotopes in appropriate box and key to Part 4.a above: Quantities Handled Microcuries Millicuries Curies Kilocuries Sealed sowas 4a. 1.2 4a. 1.2.3 4a. 1.2.3 M he. Misc. & Co-60 Misc. & Co-60 Co-60 Unseakd alpha

44. 1.2.3

44. 1.2.3 4a. 1.2.3 enarters U.Np,Pu.Th U-235. U-238 U-235 & U-238 Am Th Unuskd beta.

4a. 1.2

44. 1.2 4a. 1.2 Misc. & MAP

-Misc. & MFP. ,g MAP MAP MFP & MAP 4a. 1.4

44. 1.4.4 4a. 1.4.4 cr-a:

Neutron souras Cf-252 Cf-252 PuSe AmBe Describe prowdures simihr to those proposed in Part 2 with which you have had experience. c. Indicate months or years for each and key to Part 4.a above. 2a - 20 years. 4a 1.2 & 3 2b - 12 years. 4a 1.2 2c - 9 years. 4a 1.2 .C \\ d. Indicate which types of facilities you have used and key to Part 4.a. (X) Ordinary Chemical labort. tories 4a 1.2 (X) "Controthd Area" (Type B) laboratories 4a 2 (X) Glove boxes 4a 1.2 () Shielded glove boxes () Caves with remote manipulators (X) Field operations with portable equipment 4a 1 5. Certificace: I hereby artify that all information contained in this Statement is true and correct. f 14 El 8/ ' Signatwo of propossa user ' Date April 20,1987 Page 19

O State od California Radeloges Hesith Saccon 744 P Street Departrnent of Health sacramene. Cantorms 95814 STATEMENT OF TRAINING AND EXPERIENCE (tJse additional sheets as necesst.ry) Instruct.on: Every individual proposing to use radioacave material is required to submit a Statement of Training and Experience in duplicate to the address given above. Physicians should request Form RH 2000 when applying for human use authorizanons. 1. Narne of proposed user: John W. Harrity Position title: Test Groun faader Addren: 1825 Malden St. City: San Diego Zip: 92109 To be included on Lic. No 2468-80 in name of IRT Corocration 2. Descriptien of proposed use a. Resesrch into Rad Effects on Semiconductor Devices b. Raofation processing of materials 3. Training: L High School Graduate: Yes X No b. College or University: Name and !ccation San Diego State University 5 Years completed Degree BS/MS Course of study Physirt c. Education specifically applicable to use of radioactive material Introduction to Atomic and Nuclear Physics Nuclear Physics II Nuclear Physics Lab 4 Experience: List experience with radioactivity beginning with most recent a. (1) Dates: From June 1973 to present Title and duties: Test Group Leader - planning and supervisina test activities at various Linac. FXR. Co-60 and nuclear reactor facilities Employer, IR7 Corocration Address: P.O. Box 80817. San Dieoo. CA 92138 (2) Dates: From June 1965 to June 1973 Title and duties: Staff Physicist. conductino radiation effects studies of semi-conductor raterials and insulators Employer: Gulf Rad Tech Address: P.O. Box 81608, San Otego, CA (3) Dates: From June 1958 to June 1965 Title and duties: Physicist, assisting in radiation effects tests and studies of semiconductors and electronic circuits Employer: General Atomic Address: P.O. Rnr A160A On niagn fa AM N 87/76) l April 20,1987 Page 20 l l

rh b. Radioactive rnaterials previously used. Cite typical radioisotopes in appropriate box and key to Part 4.a above: Quantities Handled .hocuries Millicunes Curies Kilocuries 4a. 1,2,3 4a. 2,3 Scaled sources Misc. & Co-60 Co-60. Cs-137 Unsealed alpha emitters Unsealed beta. MFP 4a. 1,2,3 pmma enutters Neutron sourms - Various nuclear reactors - I c. Desmbe procedures similar to those proposed in Part 2 with which you have had experience. Indicate months or years for each and key to Part 4.a above. 1. Radiation testing of semiconductor devices and circuits 2a 4a 1,2,3 - 23 years 2. Radiation processing of semiconductor and insulating materials 4a 2,3 12 years (3) \\_/ d. Indicate which types of facilities you have used and key to Part 4.a.

4) Ordmary Chemical laboratories 4a. 1

4) " Controlled Area" (Type B) laboratorws 4a. 1.2 (4 Glove boxes 4a. 1,2 (4 Shielded glove boxes 4a. 1,2,3

() Ctves with remote ---P' *~s (x) Field operations with portable equipment 4a. 2,3 5. Cmificate: I hereby certify that all information contained in this Statement is true and correct. k) M k FI 7 sisnetw. or or p,- fem (ml %j April 20,1987 Page 21

O ~ State of California Radsologns Heeldi Section 744 P Street Department of Health Sacramento. cantorne 95814 STATEMENT OF TRAINING AND EXPERIENCE (Use additional sheets as necessr.ry) Instruction: Every individual proposing to use radioactive materialis required to submit a Statement of Training and Experience in duplicate to the address given above. Physicians should request Form RH 2000 when applying for human use authorizanons. 1. Name of proposed user: Douglas E. Willis Position title: Test Group teaser Address: 13Ef,6 Sagewood Dr. City: Poway Zip: 92064 ~ To be included on Lic. No. 2468-80 in name of IRT C6rooration 2. Description of proposed use Research and development; linear accelerator operations 3. Training: a. High School Graduate: Yes X go b. College or University: Name and location M.I.T.. Cambridae. Mass. Years completed 4 Degree B.Sc. Course of study Electrical Enaineerina c. Education specifically applicable to use of radioactive material Radiological Safety Course given by General Dynamics Corporation, Gulf General Atomic & IRT Corporation 4 Experience: List experience with radioactivity beginning with most recent a. (1) Dates: From 1973 to present Title and duties: Test group leader and Linac facility manager: testing semicondactors in radiation environment--operation 7 maintenance of Linac Employer. IRT Corporation Address: P.O. Box 80817 San Dieco, CA 92138 (2) Dates: From 1967 to April 1973 Title and duties: Test engineer & test group leader; testing electronic components, circuits & systems in radiation environments Employer: Gulf General Atomic Address: 10955 John Jay Hockins Drive. S.D. (3) Dates: From July 1960 to Oct. 1967 Title and duties Test engineer; testing electronic components & circuits in radiation environment Employer-General Atnmic Address: 10955 John Jay Hopkins Dr., 5.0. MM NA 17/78) O April 20,1987 Page 22

b b. Radioactive materials previously used. Cite typical radioisotopes in appropnate bot and key to Part 4.a above: Quantities Handled Microcurws Millicuries Curies Kilocunes 4a i Scaled sources Mjse Unsealed alpha emitters Unsealed beta. MAP (1) MAP 4 1, 2 gamma emitters (2) Au-198, Cu-64 Au-198 (3) Neutron sources c. Descibe procedures similar to those proposed in Part 2 with which you have had experience. Indicate months or years for each and key to Part 4.a above. All my experience since 1%0 has been in research and development associated with irradiating materials using Linear Accelerators, Flash X-ray machines and TRIGA type reactors. d. Indicate which types of facilities you have used and key to Part 4.a. () Ordmary Chemicallaboratories a) " Controlled Area" (Type B) laboratories (1),(2),(3) Linacs, reactors,flashX-rays () Glove boxes () Shielded giove boxes () Caves with remote manipulators () Field

• operations with portable equipment 5.

Certificate: I hereby cernfy that all information contained in this Statement is true and correct. / $agnature Of proposed user Dete i April 20,1987 Page 23 .. ~.

O State of Califorrut. Radiologic Health Section 744 P street Department of Health Sacramento. CaWornie 95814 STATEMENT OF TRAINING AND EXPERIENCE (Use additional sheets as necesst,ry) Instruction: Every individual proposing to use radioactive material is required to submit a Statement of Training and Experience in duplicate to the address giwn above. Physicians should request Form RH 2000 when applying for human use authorizanons. 1. Narne of proposed user: Jack C. Young Position title:. Linac Facility Manager Address: 1937 Rockhoff Lane , City: Escondido Zip: 92i)26 To be included on Lic. No 2468-80 in name of IRT Corocration 2. Deseiption of proposed use Research and development 3. Training: a. liigh School Graduate: Yes X No b. College or University: Name and location No. Texas State Univ., Denton. Texas Years completed 5 Degree M.A. Course of study Physics Education specifically applicable to use of radioactive material c. 1. Masters thesis utilized a Cockcroft-Walton accelerator 2. Several courses in nuclear and atomic physics 4. Experience: List experience with radioacnvity beginning with most recent a. (1) Dates: From 1573 to present Titje and duties: Princioal scientist--experimental ohysical usina a linear accelerator 252 60 Cf sources and Co sources Employers IRT Corocratio9 Address: P.O. Box 80817. San Diego. CA 92138 (2) Dates: From 1963 to 1973 Title and duties: Staff member--experimental physics using a linear accelerator, subcritical assemblies, and critical assemblies. Emplover: Gulf General Atomic Address: San Diego, CA (3) Dates: From to Title and duties: Employer: Address: RM 20044 (7/78) 9 April 20,1987 Page 24

i W b. Radioactive rnaterials previously uwd. Cite typical radioisotopes in appropriate box and key to Part 4.a above: Quantities Handled Microcurus Millicuries Curies Kilocuries Scaled sourms Co60 CsI37 60 60 Co Co Unsealed alpha UZ33 Th333 U235. Th332, emsters 238 238 U U t.'nsealed beta-32 pmma emitters P gpp Cf-252. Pu8e, , Neutron sources Pu8e. Po8e PoBe Desmbe procedures sunalar to those proposed in Part 2 with which you have had experience. c. Indicase months or years for each and key to Part 4.a above. Research and development 4a 1 - 4 years 4a 2 - 14 years s d. Indicate which types of facilities you have used and key to Part 4.a. (X) Onhaary Chemicallaboratories 4a 2 (X) " Controlled Area" (Type B) laboratories 4a 1, 2 (X) Glove boxes 4a 1, 2 (X) Shielded gion boxes 44 1, 2 () Caves with remote manipulators (X) Field operanons with parmble equipment 44 1, 2 ' 5. Certifimer: I hereby certify that allinformation contained in this Statement is true and correct. M, c. Q --4 L+ 14.195 / p sy.9 e u, w April 20,1987 Page 25

O This page Intentionally left blank. 'G April 20,1987 Pa8e 26

l V

6. ADMINISTRATIVE PROCEDURES 6.1 PROJECT AUTHORIZATION PROCEDURES Any IRT operating department or individual desiring to initiate a program utilizing SNM must first obtain the proper authorization. The following procedure I describes the proper course of action.

6.1.1 Radiation Work Authorization Request Procedures The responsible person directing the program, or his appointee, must prepare a Radiation Work Authorization (RWA) request form (see Appendix II) describing the operations to be performed, and the necessary procedures, equipment features, process characteristics, and planned precautions which assure radiological and criticality I safety. The RWA form includes the names and signatures of all persons involved in the program and identifies those persons responsible for the material specified. The I request should be reviewed with the Health Physicist before it is submitted to the RSO. This step is optional but is beneficial to the requestor. The request is then reviewed in detail by the Radiation Safety Officer who will document his analysis of the project, make recommendations, if necessary, and indicate his approval or disapproval, and if criticality safety review is applicable. I Criticality safety review shall be necessary if': 1. The RWA involves the use of 200 grams of SNM. 2. The amount of SNM used raises the inventory of a single Controlled Access Area (CAA) to 200 grams. 3. The estimated annual throughput for a single CAA exceeds 2,000 grams. Except as specifically approved by this license, the limits of approval for the 235 Committee shall be 200 grams of U. (enriched >20 percent), and 30 grams of plutonium, or 200 grams of combined HEU and Pu for laboratory-type operations with 5 p materials in any form. V Aprl! 20,1987 Page 27 h

The approval limit for use of discrete sealed sources with enrichments less than 235 10 percent shall be limited to 1400 grams of U. Where various types and enrich-ments of SNM are combined or used in these operations, the amount of material allowed will be determined by the following formula: 235 235 U (>20%) + Pu U (< 20%) # I + 200 1400 RWA's requiring criticality review will be directed to the Nuclear Systems Division Chief Scientist for additional review prior to submission to the RSC. With the approval of the RSO, the RWA is then presented to the RSC by the RSO. The RSC will review the RWA request and indicate approval or disapproval based upon specifications given in subparagraph 6.1.3 below. The RSC may approve the RWA with certain conditions or changes attached which the Committee feels is necessary for safer operation. If approved by the RSC, the RWA will be valid for :. period of one year from the month of approval and copies will be distributed as follows: Original Copy: To requestor First Copy: Radiation Safety Officer Second Copy: RWA File If disapproved, the RWA request will be returned to the initiator with the reason for rejection stated and recommendations for modification. 6.1.2 Radiation Safety Officer Review Procedures It is the responsibility of the RSO to establish that: 1. The applicant and all personnel listed in the RWA request are Authorized Individuals: each person named on an RWA must have on file with the Radiation Safety Officer a resume of his previous experience with radio-active material and/or radiation sources, training and other qualifications, to indicate competence in dealing safely with radiation and radioactive mate-rials. In addition, it is necessary that the individual has successfully April 20,1987 Page 28

C' completed an approved course in Radiological Safety, including a written examination on radiological safety principles and policies as applied to IRT activities. In some cases, due to past training and experience, a waiver of this course requirement may be granted by the Radiation Safety Committee. " Authorized Personnel" does not necessarily mean capable and competent to handle any or all radioactive materials. It merely means that these persons are authorized to be named on an RWA and either have passed a training course or have been granted a waiver. 2. Experiments which could result in an airborne concentration of radioactive material which, when averaged over 40 hours in one week, exceed the limits set forth in 10 CFR Part 20, Appendix B, Tables I and II, shall require the use of a suitable glove box or hooded facility with a filtered ventilation and, if necessary, require exposure time limitations or, in the case of emergencies, require the use of protective respiratory eq tipment. 3. Contaminated materials or experiments which could result in contamination levels greater than twice background levels must be handled in contamination V control zones, glove boxes, etc., using adequate protective clothing. In reviewing the adequacy of the facilities specified for the operation, the report entitled "Workplaces for Unsealed Radionuclides" authored by D. A. Pickler is used as a guide. Deviations from this guide are allowed if alternatives are acceptable according to the considered Judgment of the RSO and RSC. This guide is included in Appendix III. 4. Experiments involving radioactive materials with radiation levels of a magni-tude that could result in exposure levels in an uncontrolled area of: (a) 2 mrem in one hour; (b) 100 mrem in one week; or (c) 0.5 rem in one year as specified in 10 CFR 20.105 shall be done in adequately shielded cells. For controlled areas, adequate shielding and time limitations must be specified to prevent exposures in excess of those given in 10 CFR 20.101, 5. The procedures as specified in the request are presented in appropriate detail and are proper and adequate to assure safe operation. 6. All license requirements are satisfied. A r i V April 20,1987 Page 29

/ 7. The personnel involved and directing the program are competent to perform the indicated work, and have the necessary training and experience to safely use the radioactive materials authorized by this RWA. 6.1.3 Radiation Safety Committee Review Procedures In their review of the RWA request, the RSC must consider the following. 1. The adequacy of the engineering controls for the proposed project. 2. The conditions under which operations must be performed subject to the following criteria, The equipment to be used as specified in the request is adequate to a. assure safe operation. b. The required facilities have been specified and are adequate to assure safety, including the use of glove ooxes, hoods with proper ventilation systems, inert gas boxes, storage areas, containers, and shielding (experi-mental cells, structures, etc.) as appropriate. c. The procedures, as specified in the request, are presented in appropriate detail and are proper and adequate to assure safe operation. d. The personnel involved and directing the program are competent to perform the indicated work, and have the necessary training and experience to safely use the radioactive materials authorized by this RWA. For RWAs involving de use of SNM, the RSC will evaluate the proposed e. operation to assure that all reasonable precautions are taken to prevent any accidental criticality. To assist in this evaluation, the Committee has available the following publications: " Administrative Practices for Nuclear Criticality Safety," ANSI / ANS-8.19-1984, " Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors," ANSI /ANS-8.1-1983, and " Guide for Nuclear Criticality Safety in the Storage of Fissile Materials," g ANSI.ANS-8.7-1982. W April 20,1987 Page 30

The Committee's review of the RWA shall include: (1) Reliability of any engineered safeguards against criticality. (2) Familiarity of requestor with the factors related to.the occurrence of criticality. (3) Potential interaction of the proposed SNM usage with any other SNM used in the same CAA. (4). Adequacy of procedures established to prevent operational losses of SNM. (5) Adequacy of procedures to safeguard SNM. ~(6) Adherence of the proposed operation with the requirements of all applicable regulations. (7) The fire safety aspects of programs involving the use and storage of SNM in foil or finely subdivided metal particles. Approvals of work -authorizations and other actions by the Committee will be O.

V documented and kept on file for a period of one year following termination of authorization.

l In addition to the review of proposed programs, the RSC must review policies and criteria established by the Radiation Safety Organization and review operations covered by these criteria. The Committee must also audit all work subject to this license for conformance to and effectiveness of all applicable procedures and practices. The Committee will meet at least once each calendar quarter to review RWA requests and conduct other business. There must be a majority of members present to [ constitute a quorum. One of these must be the Radiation Safety Officer or his designated appointee. No RWA request may be approved or any official business transacted in the absence of a quorum. 6.2 PROJECT CHANGES AND RENEWALS RWA's are approved for a period of one year from the month of authorization and may be renewed by submittal of the RWA request 30 days prior to the termination date. Renewal without repeating the formal review requires that the RSO and/or Chief Health Physicist inspect operations and make a recommendation for renewal on the RWA request. The RSO may approve RWA renewals without formal review by the RSC. April 20,1987 Page 31 , - ~ -. - -. _ -.. _ -... - -..

I. e -i ( .-} ) t x g! x i Chlnges in a project require the filing of an amendment to the existing RWA. l The RSO"will determine the need for formal reslew by the RSC, bamd on the extent of changes since the previous review. No changes are made or renewals granted without the approval of the RSO. All changes concerning the amounts or use of SNM must be reviewed and approved by the , 1 RSC.j i 6.3 SELECTION CRITERIA FOR HEALTH PHYSICS STAFF AND RADIATION f SAFETY COMMITTEE i + ( The RSO shall have an accredited college degree in physics, the biological "i sciences, engineering, or other relevant fields and a minimum of four years of applicable experience involving radioactive materials, radiation sources, special nuclear materials, and radiation safety, or a minimum of eight years of applicable experience s involving radioactive materials, radiation machines, special nuclear material, radiation I e'" j safety, and administrative practices in th'e radiation safety field. The Health Physicist shall have a minimum of five years of applicable experience involving radioactive materials, radiatien-producing machines, special nuclear material and radiation safety. u Y Members of the R'adiation Safety Committee shall haie expertise in their fields of J specialty. For technical members, an accredited college degree in a field appropriate - to their specialty and a minimum of two years of related experience is required. For a nontechnical member, a minimum of five years of experience is required in an appropriate field. l The Radiation Safety Committee is appointed by the President of IRT and has a minimum of five members, including the RsO, a member representing'IRT management, and a member representing experimental operations. t 6.4 OPERATIONS OF THE RADIATION SAFETY OFFICER AND HEALTH PHYSICIST } In addition tar the above-mentioned responsibilities, it is the responsibility of the [ Radiation Safety Officer to: ) Review anb inspect all operations at least yearly and more frequently if l. specified on an RWA. s Apr;l 20,1987 Pa 1 > ge 32 j i

D 'V- ' 2.. To inspect and assure proper operation of all new installations affecting . materials governed by this license before accepting them as workable and usable. It is the responsibility of the Health Physicist to: .l. Initially investigate all incidents involving radioactive materials, special nuclear materials and radiation sources to determine the causes and circum-stances and what corrective action is needed, and report findings to the RSO and RSC. 2. Ensure that policies of the RSO and special requirements as specified on the RWA's are being followed and to close down all operations that are violating these policies or RWA requirements until the situation is remedied. 3. See that radioactive waste material is handled properly,' that the proper containers are available and being used, and that all waste material is being l disposed of in a proper and timely manner. l It is the joint responsibility of the Radiation Safety Officer and the Health Physicist to: - 1. Maintain a complete centrally located file of all records required by this license. 2. Investigate the design criteria of all facilities and ensure that they are functional for their intended use and radiologically safe. 3. Take charge of all radiological emergencies and institute remedial action as provided by the general emergency procedures. l 6.5 INTERNALINSPECTION AND REVIEW j The Radiation Safety Officer is ' responsible for inspection and review of all activities involving materials subject to this license to establish that all activities are authorized, that they are in compliance with this license, and that good radiological protection practices are being used. He is required to submit a written report of any discrepancies to the Radiation Safety Committee, the cognizant IRT Department Manager, and the President of IRT. Lo April 20,1987 l. Page 33 ._,,._s

These audits are performed on an annual basis unless specified hazards require more frequent auditing. The frequency of auditing of projects is established by Radiation Safety Committee and included (written) in the RWA. 6.6 OPERATIONS OF THE RADIATION SAFETY COMMITTEE In addition to its authorizational duties described above, the Radiation Safety Committee must: 1. Review :nd establish organizational policies and procedures concerning radiological matters and make recommendations to management for their implementation. 2. Review and investigate all significant incidents and report to management with recommendations as to remedial action and policy or procedural changes if indicated. 3. Assure that all operational difficulties such as unsafe practices, noncompil-ance with license, safety hazards, etc., are corrected prior to continuance of program or procedures. 6.7 CONTROL AND ACCOUNTABILITY OF SPECIAL NUCLEAR MATERIAL 6.7.1 Organization for Control, Safeguarding, and Accountability of SNM Inventory control of SNM will be accomplished through a system of Permanent and Temporary Controlled Access Areas (CAA's) which are designed to minimize interaction of separate uses and provide a basis for accountability. Each CAA will be a well-identified physical area with specified boundaries. Each CAA will have a Material Custodian and Alternate Material Custodian who will be responsible for accountability and adherence to license criteria. The operating and review functions for' material control, accountability, and prevention of criticality will be conducted by a Nuclear Materials Manager (NMM) and the Radiation and Criticality Safety Committee (RSC). 1. Responsibilities of Nuclear Materials Manager The Radiation Safety Officer is the Nuclear Materials Manager (NMM). The E NMM will have overall responsibility for operation of the Control and Accountability System. The responsibility includes: April 20,1987 Page 34

D' a.. Assurance that all receipts and shipments are within license limitations. b. Maintain material control and accountability procedures. c. Report any losses as required by applicable regulations.

d.. Maintain a file of material transfers and current inventory for each MBA.

e. Act as Material Custodian for the fuel storage areas and control access to these areas, f. Perform a physical inventory of all CAA's annually. g. Maintaining adherence to IRT Corporation's approved Physical Security Plan. l 6.8 STORAGE OF SPECIAL NUCLEAR MATERIALS 6.8.1 3030 callan Road Storage Vault This vault-type room is located on the first floor, North wing of the new IRT { Headquarters Building. The walls and ceiling are made of 8-inch thick reinforced concrete block. A metal door equipped with a combination-type security padlock provides entrance. The vault shall be equipped with an ultrasonic motion monitor for security. The room measures 6 feet by 6 feet by 8 feet high. Inside the fault, immediately opposite the entrance, are two 4-inch diameter I pipes that extend 16 feet under the parking lot, The pipes are 6 to 7 feet underground. The buried end is welded shut and the end in the vault is closed with a pipe cap. These tubes will be used to store long reactor test rods. The other storage compartments will consist of metal cabinets measuring 18 inches deep by 36 inches wide and 72 inches high. Storage bins shall be built on each shelf in order to comply with the Storage Criteria discussed in Item 3, which follows. 6.8.2 Storage Criteria The rules for storage are governed by both density and area criteria with the following rules governing storage in the HTGR vault. 1. Unmoderated SNM metal, alloy or compounds may be stored in closed 235 containers limited to 1.2 kg of 0 in a 3.6 liter volume,0.8 kg of plutonium 233 d in a 2.4 liter volume, or 0.4 kg of 0 in a 1.3 liter volume. April 20,1987 Page 35

2. Moderated SNM (H/X greater than two) may be stored in isolated plane 235, 0.098 kg of plutonium, arrays not to exceed an average of 0.160 kg of U 233 or 0.140 kg of U per foot square of aspect area. No foot square of the 235 aspect area may contain more than 0.320 kg of U, 0.196 kg of plutonium, or 0.280 kg of 233U The " foot square of aspect area" means any foot square area as viewed. There are two restrictions governing the storage, (1) an average loading of x grams per square foot, and (2) 22 grams in any one square foot of aspect area. Uniform loading of x grams per square foot over an entire array satisfies the storage criteria--no more than x grams per foot square regardless of where the square foot area is selected. The array shown in Figure 1 also satisfies the storage criteria. Some of the material is concentrated in a smaller area; however, no single aspect area is allowed an excess of 2 x grams of material. Referring to the figure if each box defines a foot square there can be x grams on the average in each square. You can also store x grams in adjacent corners so long as you do not exceed 2 x in any foot square of aspect area, shown here as the dashed foot square. Any material stored in the other two connecting squares must fall outside the dashed foot square. R R p.._q !R I I L..J R R R I I RT-20982 Figure 1. Fuel storage array depicting aspect area There is a further degree of safety in the storage of SNM because the storage racks and compartments are made of boral. The basic reference for critical dimensions of SNM-moderated systems is 235 TID-7028; H. C. Paxton, et al., " Critical Dimensions of Systems Containing U, April 20,1987 Page 36

n 1 239 233 Pu, and U," USAEC Report TID-7028, Los Alamos Scientific Laboratory and Oak Ridge National Laboratory, June 1964. This document is an extensive cross comparison of many critical experiments on a common basis. Rule (1) above is based on the critical parameter versus SNM density curves, 5 Figures 8, 9, 27, 28, 34, and 35 of TID-7028, with a safety factor of 2.3, which provides double-batch protection under mass control and 1.33 in volume. These curves are for 4 water-moderated SNM spheres. The metal-water mixture data have been used since these figures, in nuclear safety evaluation by the licensee, assume the SNM exists in the containment vessel as a fine particulate material and the vessel may become water flooded. Reflectors, other than water, have been taken into account by use of Table 7 and Figure 44 of-TID-7028. The most effective moderator is Be,. extrapolated from Figure 44 to be 7.5 kg for an infinite reflector. However, Be is not a common reflector and it is very unlikely that a storage situation would arise in which the entire container i would be reflected by Be since the shelf umts are commonly made of Boral. Using the - data from Table 7 and Figure 44 of TID-7028, reflectors other than water have been i taken into account by further reducing the critical parameter versus SNM density curves using another safety factor. This safety factor is determined by the ratio of the l critical mass of an infinite beryllium reflector to the critical mass for an infinite water 1 reflector for an'unmoderated SNM sphere. This safety factor is 0.33, and although it 235, it has been used for plutonium and 233, applies directly to U g The nuclear safety of this geometry is based on the fact that 1.2 kg of '35U is 235 safe in a spherical shape for all metal-moderator mixtures above 1.7 kg U per liter and 3.6 liters is a safe spherical volume for any metal-moderator mixture below 1.7 kg 2350 per liter. 235 A similar argument to that for U applies for plutonium and 0.8 kg of plutonium. is safe in a spherical shape for all metal-moderator mixtures above 1.4 kg plutonium per liter and 2.4 liters is a safe spherical volume for any metal-moderator mixture below 1.9 kg plutonium per liter. 233 [ Similarly, the U case is safe, based on the fact that 0.4 kg is safe in a 233 spherical shape for all metal-moderator mixtures above 0.96 kg 0 per liter and 233 1.3 liters is a safe spherical volume for any metal-moderator mixture below 3.9 kg g l per liter. lO April 20,1987 [ Page 37 i l 4 w ~ r e -a - y r_- e,-. w rv, mw ww -e--m p -n,g,-er,----e.--+mmm-,- ,4., ,.,-9, ,,,,,e,m,,e-e, ew_m.--w w,,-,.w ,,w, m aw w., m. w-,n m ww n mm m e ac- -

Moderators other than water, such as carbon and beryllium, as a diluent, have less moderation ability, per atom, than hydrogen and the addition of carbon and beryllium result in a loss of reactivity. Rule (1) is also based on Table IV and V of TID-7016, Revision 1, for plane lattices of this type. This is, therefore, based on a minimum center-to-center spacing of 16 inches and surface-to-surface spacing of eight inches or greater. Rule (2) is based on the slab geometry curves of TID-7028. The area-density concept is of frequent use in storing and handling SNM; this is usually expressed in kilograms of SNM per foot square of area. The working limits are arrived at by using water-reflected critical slab geometry curves of. TID-7028, converting to equivalent kilograms per foot square, applying a safety factor of 2.3 for double-batch protection, and replotting versus the SNM density per liter of SNM-water mixture. The minimum 235 of these curves are at values of 0.160 kg of U, 0.098 kg of plutonium, and 0.140 kg 233 of U per foot square. 6.8.3 Arjons Warehouse Storage An NRC-approved shipping container will be utilized as the storage container for the Plutonium-Uranium test rods. The container is: Certificate of Compliance No. 6581, Model 51032-1. The rods will be stored inside two wooden inner containers, then placed inside the shipping container, and the shipping container will be inside a locked storerocm. The wooden boxes are made of 1-inch stair tread stock that is glued and nailed together, except for the top which is fastened on with 2-inch wood screws. The inside dimensions of the box are 4.125" x 4.125" x 76.75"(see Figure 2). The shipping / storage container is a 43 inch diameter right cylinder 216 inches long fabricated of 3/8-inch steel. This vessel is fabricated in two sections, a base and a cover assembly. Continuous 2 x 2 x 1/4-inch closure flanges are welded to the base and cover assemblies. A 1/2-inch continuous rubber "O" ring gasket is fitted between the mating flanges. The two halves of the containment vessel are mated and sealed together with fif ty-eight 1/2-inch 13UNC-2A steel closure bolts; in addition, a padlock and a tamper indicating seal will be attached to each end of the cask. The complete assembly weighs about 4000 pounds (see Figure 3). The storeroom is 32 feet long,12 feet wide, with a 20 foot high ceiling. 'Ihe door will be locked and is equipped with intrusion alarm contacts. There is an ultrasonic 5 April 20,1987 Page 38

k motion monitor inside the room. The alarm system will be "ON" at all times except I when it is necessary to enter, open the storage container and inspect the rods. The building has perimeter alarms on all of the entrances and there are infrared motion detectors throughout the inside of the building. This intrusion alarm system is u ON" at all times, except when the building is occupied (see Figure 4). 6.8.4 Temporary Storage of SNM in Use The temporary storage of SNM in active use will be accomplished by the use of locked metal cabinets located in the temporary CAA. Access will be controlled by keys in the possession of the Material Custodian of this CAA. Physical security will be in accordance with IRT Corporation's approved Physical Security Plan. 6.8.5 General Procedures for Use of Special Nuclear Material Several general procedures apply to all users of SNM. 1. The Criticality Safety Committee must approve all operations which fall (n within the ilmits on SNM specified in Section 6.1.1. 2. All transfers of SNM between CAA's must be recorded on a Material Transfer form. One copy must be given to the NMM and one retained by the Material Custodian. 3. Each Material Custodian must maintain a log of transfers in and out of his CAA, showing a running inventory of the current amount in his possession. 4. Unless specifically exempted by the License, a criticality monitor which conforms to 10 CFR 70.24(a)(1) will be operating in each area where more 235, 300 grams of Pu, or 300 grams of 233 than 500 grams of contained U U are present. Special nuclear material operations are generally controlled by mass limits. The CAA is prevented from exceeding its limits by the Nuclear Material Manager controlling SNM issuance to the CAA. The CAA log book for the operation provides the means of enforcing and auditing adherence to these limits. The possession limit under each CAA is the mass limit authorized and periodic audit ensures that the CAA books are properly maintained, that limits are not exceeded, and that operations are safely conducted. Aprl! 20,1987 Page 39 I

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m D) a' In addition, the entries leading into a CAA shall be conspicuously posted by a sign or signs. Such signs shallinclude the following information: a. SNM limitation b. Quantity of SNM currently in the CAA c.- Material Custodian d. Alternate Material Custodian (s). The quantity of SNM within the CAA will be updated whenever SNM is added or removed. When the Material Custodian or alternate Material Custodian is not in the immediate vicinity of the CAA, entries leading 'into the CAA will be controlled by locked doors, a watchman, or a security guard for authorized personnel entry. Anyone who wilfully violates the above administrative controls will be subject to appropriate corrective action. Any future RWA's signed by the violator will have a notation beside his name as having " violated a good radiation safety practice." The cognizant manager of the violator will be notified in writing of the violation and a record of the violation will become part of the employee's radiation record. O Q In the event of repeated violations, the employee will be subject to disciplinary action such as denial of use of SNM, suspension or termination. 6.9 PROGRAM RECORDS The following records are maintained in a centralized file of the Radiation Safety Office. 1. Radiation Work Authorizations that include location of use; names of personnel involved; description of the program, procedures, materials, and facilities; results or comments of reviewers. 2. A list of all authorized personnel. 3. The results of all inspections and audits of the program, including survey and compliance data. 4. SNM accountability records, including material transfers, physical inven-tories, and material status reports. 5. Exposure and bloassay histories of all users. April 20,1987 Page 43

6. Leak test and environmental survey records within and outside controlled i areas. i 7. Histories of all incidents and unusual occurrences. 8. A complete file of calibration data on all instruments used for radiation level monitoring. 9. Resume of each authorized person's training and experience with radioactive materials. 9 O Aprl! 20,1987 Page 44

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7. RADIATION PROTECTION PROCEDURES 7.1 RADIATION PROTECTION MANUAL The manual that is used for instruction of personnel in the principles of radiation safety and protection is the "IRT Corporation Radiological Safety Guide" prepared by P. R. Maschka and K. L. Crosbie, as revised August 1986. In addition, the following publications are available for use in the instruction programs:

1. " Introduction to Radiological Health," by Hanson Blatz, McGraw-Hill,1964.

2. " Accelerator Health Physics," by H. Wade Patterson and Ralph H. Thomas, Academic Press,1973.

3. " Radiological Safety Aspects of the Operation of Electron Linear Accelera-tors," by William P. Swanson, IAEA,1979.

G

4. " Atomic Radiation" Theory, Biological Hazards, Safety Measures, and Treat-ment of injury, pub!!shed by RCA Service Company.

5. " Radiological Health Handbook," published by the U. S. Department of Health and Human Services, Public Health Service,1970.

In addition, we have a large number of reports, booklets, and pamphlets from the IAEA, NCRP, and NBS. These documents are maintained by the Radiation Safety Office to provide current and concise source books covering radiological safety practices and procedures estabilshed by the licensee. These books, in addition to describing the complete radiological safety program, cover the procedural needs for establishing an activity involving radioactive materials, the personnel work rules which must be employed, the kind of personnel monitoring which will be used, records requirements, etc. These manuals specify the need for special review of untried operations before commencing activities. They also identify various detection programs, such as the bloassay program designed to detect and measure radiation exposures from internally deposited radio-active materials. They are available to each individual who expects to utilize and be t April 20,1987 Page 45

responsible for radioactive material or radiation-producing devices, and this individual is charged with the responsiblity of seeing that personnel working under his direction abide by the rules and regulations. The individual user at all times is further charged not only with the responsibility of his own safety but also with the safety of others and the protection of equipment and facilities. 1 Changes to the IRT Radiological Safety Regulations can only be made by the Radiation Safety Officer, with the approval of the Radiation Safety Committee. 7.2 PERSONNEL MONITORING 1 In order to fulfill responsibilities for evaluating and recording all exposures that are actually incurred, personnel monitoring equipment is provided for each employee who may be subject to irradiation and any visitor who is permitted into a radiation area. Appropriate personnel monitoring equipment shall be supplied to and be required for use by: 1. Each individual who enters a restricted area under such circumstances that t he receives, or is likely to receive, a dose in any calendar quarter in excess of 25 percent of the applicable value specified in 10 CFR 20.101(a). I 2. Each individual under 18 years of age who enters a restricted area under such circumstances that he receives, or is likely to receive, a dose in any calendar quarter in excess of five percent of the applicable value specified in l 10 CFR 20.101(a). 3. Each individual who enters a high radiation area. 7.2.1 Film Badges Film badges are used to measure external radiation dose for occupational l personnel. The following table gives a summary of the typical characteristics of such I devices. Self-reading pocket lonization chambers (dosimeters) are used in addition to film badges as determined by licensee to measure x-ray and gamma radiation, fast neutron radiation, and thermal-neutron radiation from a dose of 1 mrem to 600 rems. O April 20,1987 Page 46

/N. .C). Film Badge Dose Ranges Radiation - From To Type (mrems) (rems) Energy X-ray and gamma 10 500 5 kev to 100 MeV (high energy) X-ray and gamma 10 60 Low energies Beta 40 1000 400 kev to 50 MeV Thermal neutrons 10 300 0 to 3.5 eV Fast neutrons 15 10 5 kev to over 100 MeV The film badge and finger ring dosimetry service is provided by contract with the R. S. Landauer Company, a commercial supplier of this service. l Film badges and finger rings are processed monthly by the supplier, and pocket dosimeters are read weekly and more frequently if exposure of the wearer is likely to exceed 100 mrem during the weekly period. I Personnel-monitoring records are confidential information, but each employee or his supervisor may review his exposure record upon request. Doses in excess of 100 mrem / week are called to the attention of an Individual and his supervisor immediately following the reading revealing such an exposure, and an inquiry is made regarding the cause with an explanation noted in the record. In addition to the personnel monitors, a bloassay program for employees working with unencapsulated radioactive materials is conducted, utilizing urinalysis, lung burden assay, and whole-body counting as appropriate to provide the company with a method i for acquiring records related to occupational radiation exposures. Bloassay services are contracted to United States Testing Company, Richland, Washington, while whole-body counting and lung burden assay services are available from the General Atomic Health Physics Department. The program is as described below, f 7.2.2 Bionssay l Bloassays will be conducted in accordance with the guidelines contained in l Regulatory Guide 8.11. f O April 20,1987 Page 47 l i , _, ~ - _. _

Those individuals working with ingestible radionuclides are required to routinely submit urine specimens for analysis. The analysis and interpretation of these bioassays are done according to ICRP Publication 10. Each individual working with ingestible radionuclides is assigned a hazard index number of one to five based on his own situation. The exposure hazard indices for gross alpha bloassay are shown in the following table. In regard to the sensitivity of the analysis, the present minimum sensitivity of the bloassay analysis is <1 percent of the maximum permissible body burden for a specimen void internal of eight hours for uranium and tritium and <10 percent for other time intervals and materials. For whole-body counting,.the minimum sensitivity is less than 1 percent maximum permissible body burden for each nuclide. Exposure Hazard Indices for Gross Alpha Bioassay Exposure Hazard Sample Typical Index Frequency Situation 5 Weekly Special schedule for individuals whose average bioassay results exceed 4.0 pCi/ day of alpha l activity, or who worked in an area where the airborne activity is >25 percent of DAC. 4 Monthly Individuals working in areas where the quar-terly airborne alpha activity is >10 percent of DAC and <25 percent of DAC. 3 Semiannually Individuals working in areas where the quar-terly airborne alpha activity is <10 percent of DAC and the maximum used to obtain the average is <25 percent of DAC. 2 Annually Individuals working in areas where the quar-terly airborne alpha activity is >l percent of DAC and the maximum used to obtain the average is <10 percent of DAC. 1 . As needed Individuals who work in potential alpha con-taminated areas or following an accidental release of airborne alpha emitters into the work area. O April 20,1987 Page 48

(M () The exposure hazard indices for tritium are shown in the following table. F===re Hazard Indices for Tritium Bloassay Exposure Hazard Sample Typical Index Frequency Situation 5 Next day Work with 10 Cl or more of uncontained tritium 4 Monthly Robtine operations involving 0.1 Ci or more of tritium 7.2.3 hie-Body Monitoring Whole-body monitoring is routinely done on all individuals who may become internally contaminated with beta or gamma emitting radioactive materials. Each individual working directly with radioactive materials and each individual who has a reasonable probability of becoming contaminated if an incident should occur is assigned a hazard index number of one to three based on his own situation. The exposure hazard indices for whole-body monitoring are shown in the following table. Fm=re Hazard Indices for hie-Body Monitoring Exposure Hazard Sample Typical Index Frequency Situation 3 2 or more Individuals who show continuing or recurring times a year evidence of internal contamination. 2 Annually Individuals who routinely work with more than 10 times the MPBB of any unsealed alpha emitter. 1 As needed All individuals who work with uncontained SNM in an ingestible form Additions, deletions, and changes to these programs or procedures are made by the Health Physicist. O April 20,1987 Page 49

7.3 LIMITS OF RADIATION IN CONTROLLED AND UNCONTROLLED AREAS 7.3.1 Controlled Area Controlled areas include any area, whether the property of IRT or temporary job sites, in which the company controls access for purposes of radiation safety. Access is controlled by means of fences, ropes, or other barriers which deter access by the general public. Areas which have radiation levels equal to or greater than that which will cause exposure of (a) 2 mrem in one hour, (b) 100 mrem in one week, or (c) 0.5 rem in one year or have smearable contamination levels greater than (a) 140 disintegrations per minute per 100 cm2 above background, of beta-gamma emitters, or (b) 26 disinte-grations per minute per 100 cm2 above background of alpha emitters are classified as controlled areas. Areas with radiation levels of 5 mrem per hour are declared radiation areas and are monitored and posted with appropriate signs by the Health Physicist or the individual responsible for the program. Areas in which the airborne radioactivity exists in concentrations greater than those specified in 10 CFR 20.203, or which, averaged over the number of hours in any week during which individuals are in the area, exceed 25 percent of the above concentrations, are declared airborne radioactive areas and are monitored and posted as such by the Health Physicist. Work in these areas is done only with appropriate clothing and respiratory protective devices. Radioactive Work Permits (RWP) signed by the Health Physicist or the RSO are necessary for all work in controlled areas for personnel who are not covered by an RWA or who have not completed the Radiological Safety Course given by the company and do not have on file with the RSO a resume of previous training and experience and/or a waver. 7.3.2 Uncontrolled Areas Uncontrolled areas are those areas wherein the company does not control access. Radiation levels in these areas shall be as low as practicable, but will not result in exposure of more than (a) 2 mrem in one hour, (b) 100 mrem in one week, or (c) 0.5 rem in one year at the boundary of the controlled area. Release limits of effluents, either airborne or as liquids, are controlled to have an average annual activity at the boundaries as far as practicable below the maximum permissible concentrations (MPC) for unrestricted areas as defined in 10 CFR 20. Any April 20,1987 l Page 50 1

O time the accumulated data indicate that a level of 50 percent of the MPC may be exceeded on an annual basis, corrective action is instituted. The discharge of radioactive material into the sanitary sewer system will be in compliance with 10 CFR 20.303 and will not exceed 25 percent of the limits specified. ~ The release of decontaminated facilities and equipment shall be within the guidelines set forth in: ANNEX A " Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct, Source, or Special Nuclear Material." In addition, every reasonable effort shall be made to reduce residual contamination levels to as low as practical. 7.3.3 Contamination Control Surface contamination is controlled by containment. Cross contamination is limited by employing practical hand!Ing techniques. When routine contamination control falls, decontamination procedures are used. A reading of twice background is v readily detectable and is accepted as indicating positive contamination. Surface contamination is detected and evaluated using: (a) large area wipes checked with a survey meter;(b) direct survey; and (c) wipes measured in the IRT Health Physics Lab. Airborne contamination is controlled by containment in fume hoods, glove boxes, facility ventilation systems, chemistry hoods, and high-efficiency air filters. The techniques used for contamination measurement are detailed below. Filtered exhaust systems throughout this application imply a system containing high efficiency particulate filters (99.7 percent efficient in removing particles as sniall as 0.3 micron in diameter). Typically these filters are preceded by Dust Stop pre-filters. All work with special nuclear materials, in either dry or liquid form, which could create airborne radioactivity will be done in a glove box or chemistry hood equipped with high-efficiency filtered exhaust systems and a means for collecting any liquid waste material. E The pre-filters in the chemistry hoods and glove boxes will be changed when there has been a throughput of 1 kg of unsealed SNM through the particular device. O i' April 20,1987 Page 51

The exhaust streams of these systems, when in use for special nuclear material, will be continuously sampled by portable air samplers located downstream of the high-efficiency filters. The sampler head is centrally located within the exhaust stream just prior to the exit such that the air stream is sampled as it enters the atmosphere. The portable air sampler utilizes a Gast Model 0521-V3 or equivalent Carbon Vane, constant volume, vacuum pump with a Rockwell 310 or equivalent gas flowmeter. l Air samplers have their filters changed each work day. The filters are counted and evaluated by the IRT Health Physicist. The samples are routinely counted and evaluated at least twice. The first count takes place as soon as possible after the sample is collected. This gross count gives an indication of the immediate condition in the facility. The second delayed count and, in some cases, a third delayed count enable a more quantitative measurement of the long-lived nuclides present in the sample. Any time the accumulated data indicates that a level of 20 percent of MPC may be exceeded on a weekly basis, corrective action shall be instituted. More frequent sample changes are accomp!!shed and other analysis routines are initiated as required during nonroutine operations or accident situations. Internal personnel radiation exposure is evaluated through the bloassay program and total body counting. 7.3.4 Radiological Survey Radiological surveys will be conducted in order to determine the radiological hazards involved in new and untried operations and to ensure that operations in process remain under proper controls. The frequency and type of surveys are determined by the Health Physicist, the RSO, principal investigators, and the RSC. The basis considered l are type of operation, hazards involved, and experimental conditions. These surveys are made at various frequencies, depending upon the conditions. They may vary from as frequently as several times per shif t (every entry into a cell) for Linac experiments, to once per year as with surveys of the office areas. Air samples are taken in each CAA where unsealed SNM is used with the sample head positioned as'near to the workers' breathing zone as is practicable. Using the guidelines outlined in Appendix 111 of this app!! cation, the work is conducted in chemistry hoods or glove boxes as is necessary. As a matter of general practice, all work with powdered or liquid SNM which is in a readily ingestible or respirable form is done in 9 April 20,1987 Page 52

/^% either a hood or glove box. As a result of this policy, air samples taken in the work areas during previous operations have shown a quarterly average concentration of airborne radioactivity of: 1.1 x 10'I3 pCl/cc for all areas, with 1.6 x 10-I3 Ci/cc being the highest, and 3.8 x 10~I" C1/cc being the lowest concentration for individual areas. 7.3.5 Off-Site Operations Off-site operations involving SNM covered by this license are accomplished under the provisions of this license and applicable procedures for project authorization, radiation 1.rotection, and safeguards. Notification of off-site work will be made using an NRC-241 form. Authorization requests contain the necessary information regarding the location, materials involved, the establishment of controlled areas, safeguard procedures, and adequate shielding such that the radiation levels at the boundaries do not exceed those specified in 10 CFR 20.105 and 20.106. All material handling and control of material is accomplished directly by or under the direct supervision of IRT-authorized individuals. All necessary radiation monitoring devices and alarm systems are used and posting and labeling requirements specified in 10 CFR 20.203 are adhered to. It is the direct responsibility of the IRT principal investigator or off-site supervisor to assure safe operation and physical security of licensed material. He also assumes responsibility for all site monitoring and personnel monitoring activities. In discharging the above responsibilities the principal investigator or off-site supervisor establishes boundarles of the controlled access area and area controlled for purposes of radiation safety. Barriers are erected as appropriate to exclude personnel not under his control. He provides film badges and dosimeters to any personnel involved who are not IRT employees and is responsible for instructing them on procedures and precautions to be followed. All personnel involved are directly under this control, including those who are not IRT employees. If the facility is also a licensee, all activities are coordinated with the facility RSO and joint responsibilities are agreed upon. O April 20,1987 Page 53 i i

It will be the responsibility of the RSO or the HP to indoctrinate all field personnel in the particular radiation safety problems associated with their off-site project, and to instruct them in the proper radiation safety procedures with emphasis on radiation monitoring, personnel monitoring, and radiation protection. At the conclusion of operations, the principal investigator or off-site supervisor is responsible for safe removal of all radioactive material, together with a thorough survey of the site to assure that it is clear and free of contamination prior to removing his control of the area. Transportation of radioactive material to and from the off-site locations is in accordance with the current NRC and DOT regulations, and IRT Corporation's approved Physical Security Plan. O t O April 20,1987 Page 54

,cy (v) S. INSTRUCTION OF PERSONNEL 8.1 FORMAL TRAINING TOR NEW USERS IRT conducts a program for Instructing new personnel regarding health and safety rules and problems attendant to the use of sources 'of radiation. The radiation protection manual has been described in Section 7. An outline of the Radiological Safety Orientation Course based on the IRT Radiological Safety Manual, IRT 4171-009, is included as Appendix IV. The course is supplemented with I study handouts and terminated with an examination; a grade of 70 is required for passing. All personnel currently working for IRT, and who are also working unsuper-vised with radioactive material, have passed the Radiological Safety Test or have been granted a walver. All new personnel who will work with radioactive material or within a controlled radiation area are required to take the Radiological Safety Orientation Course. 8.2 PERIODIC RETRAINING At least every two years a short review course is given to all personnel working with radioactive materials. April 20,1987 Page55

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9. TECHNICAL CAPABILITIES 9.1' GENERAL PURPOSES OF USE IRT Corporation's programs which utilize radioactive materials and radiation producing machines cover three major categories--instrumentation, research, and technology. These are further subdivided into nuclear energy, security and safety, non-destructive testing, resource exploration, survivability, radiation effects, and radiation services. The programs range from basic research to finished products.

Instrumentation programs cover four major areas (1) resource exploration, (2) nuclear material measurement, (3) nondestructive testing, and (4) data acquisition and analysis systems. Instruments include uranium borehole loggers, geothermal well loggers, fuel rod scanners, nuclear assay systems, portal safeguards monitors, contamin-ation monitors, fixed and portable neutron radiography systems, radiation gauges, ordnance quality control systems, large-scale field experimental systems, environ-mental measurement systems, and laboratory data collection and reduction systems. Research areas of interest are: (1) solid-state physics, (2) nuclear physics, and (3) electromagnetics. Experimental and analyticalinvestigations programs cover a wide variety of subjects including optical detectors, semiconductor materials and devices, insulator materials, radiation transport, reactor neutronics, activation analysis, electro-magnetic signal generation, propagation, coupling, and interference phenomena. Technology development and transfer programs cover the broad areas of: (1) electronic and electro-optical systems, (2) nuclear fuel cycle and alternate energy sources, (3) radiation processing, and (4) defense analysis. Programs involve test and evaluation, radiation and electromagnetic effects, uranium exploration, process devel-opment, material safeguards, environmental assessments, advanced fuels, geothermal sources, radiation sterilization, radiation enhancement, target camouflage, communica-tions, system vulernability, and nuclear survivability. 9.2 ORGANIZATIONAL STRUCTURE IRT Corporation's organization chart is shown in Figure 2. The two technical O division managers are responsible to the President of IRT for all matters relating to the April 20,1987 Page 57

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. technical excellence of work performed by their-respective organizations. They . establish business goals in their assigned technological areas and are responsible for the appilcation of all resources and personnel necessary to accomplish those goals. The engineering and production requirements of the divisions are satisfied by the Engineer-Ing and Manufacturing organization within the Nuclear Systems Division which has the major portion of the hardware programs. Project organizations are established for major projects cutting across disciplinary lines. The projects draw from the technical organization for both project management and technical personnel. Projects involving the use of radioactive materials and/or radiation-producing machines are reviewed by the Radiation Safety Officer and Radiation Safety Committee to assure adequate facilities, procedural requirements, and licensing authority are provided. Support for the projects and divisions is provided by the other divisions and departments listed on the organizational chart. Matters pertaining to Security and Safety are directed through the Human Resources Department. Health Physics reports I to the Radiation Safety Officer. The function of the indicated health and safety . personnel and review committees have been discussed in an earlier section. Within this structure, the responsibility for establishing and maintaining safe operations rests O-directly with the line management of the organization; and to assure the discharge of this responsibility, the President of IRT Corporation appoints review committees as appropriate. These committees are independent of management control in performing reviews and report items requiring action to the appropriate level of management. If satisfactory action is not taken, as determined by subsequent audit, then higher levels of management are advised of the need for action. The final authority on the response of management to recommendations of the review committees rests with the President - of IRT Corporation. 9.3 TECHNICAL PERSONNEL The technical staff members comprise approximately 163 of the 241 employees of the company. Of the 163 technical staff member,13 percent hold doctorate degrees, 19 percent hold masters degrees, and 31 percent hold bachelors degrees. Of the remaining 29 percent, about 30 percent have associates degrees. The academic disciplines represented by these degrees are predominantly physics and engineering,41 and 40 percent, respectively. The remaining 19 percent are divided among mathe-matics, computer sciences, chemistry, and other disciplines. The technical staff members are grouped in organization levels as shown in the following table. April 20,1987 Page 59

\\ l e, s \\ i m -J. Organizational i Level Title t i I 0 President, Vice President i 1 Research Advisor 2 Principal Scientist Principal Engineer 3 Staff Scientist Staff Engineer 4 Senior Scientist Senior Engineer 5 Engineer Physicist Technical Specialist 1 Personnel in organizational levels 1 and 2 are responsible for planning and managir;g research programs and generally have an advanced degree with an average g experience of 24 years. Personnel in organization level 3 have an average of 18 years experience. Average - Percentage of Total Technical Personnel Years Level Ph.D. M.S. B.S. None ~ Experience 0 50 50 23 1 67 34 27 2 50 30 20 25 3 10 30 50 10 20 4 20 35 45 15 5 26 74 10 R6sumis of key personnel are shown on the following pages. O April 20,1987 Page 60

e MICHAEL S. BERNATH B.S.E.E., University of Missouri; 3.D., St. Lools Unive sity School of Law; Admitted to St. Louis, Missouri, and Federal Bars Business Experience 1967-Present GOULD INC. Electronics Division 1984-Present Group Vice President Industrial Automation / President and General Manager Programmable Controls Andover, Massachusetts 1983-1984 President and General Manager Systems Protection Division Philadelphia, Pennsylvania 1980-1983 President and General Manager Electrical Components Division Long Island, New York 1978-1980 Vice President and Director of ( Marketing Development Distribution & Controls Division Chicago, Illinois 1977-1978 Vice President and Director of Marketing Electric Motor Division (Century Electric) St. Louis, Missouri i 1970-1977 Electric Motor Division St. Louis, Missouri Marketing Manager Regional Sales Manager Various Product Manager Positions Various Application Engineering Roles Various Operating Assignments. OV April 20,1987 Page 61

W. DENNIS SWIFT B.S.,, M.S., and Ph.D., Electrical Engineering, Iowa State University Dr. Swift, who joined the IRT staff in 1972, is presently Vice President and Manager cf the Electronic Systems Division. He has overall responsibility for radiation effects assessment and hardening programs, including EMP, IEMP, SGEMP, and TREE, as well as IRT's linear accelerator (Linac) facility. Dr. Swif t coordinates the activities of the Electronic Systems Division with those of the rest of IRT to ensure that the broad interdisciplinary skills within IRT are effectively utilized on programs. Dr. Swift's previous assignment at IRT was as Manager of the Electromagnetic Effects Department where he was instrumental in the development of several analytical and experimental techniques for assessing the effects of EMC/EMP/IEMP/SGEMP on complex systems. The analytical techniques Dr. Swift has developed include several EM interaction codes, including codes utilizing the method of moments. IRT's fast rise-time and high level EMP testing and cable shield testing capabilities were developed under his direction. Dr. Swift has been involved in EMP testing, current injection testing (both for structural excitation and for black box testing), SGEMP tests using both bremsstrahlung and exploding wire photon sources, and underground nuclear tests. For three years prior to joining the company, Dr. Swift served as Instructor in Electrical Engineering at Iowa State University. While at Iowa State, he taught courses in electromagnetic field theory and conducted research on the numerical analysis of antennas. Dr. Swif t was employed by the Boeing Company from 1968 to 1969, where he worked on several inertial navigation systems. While at Boeing, he developed an error analysis code for determining the attitude errors caused by various computational algorithms used in strapdown inertial reference systems. While a graduate student from 1966 to 1968, Dr. Swift designed electronic and optical instruments in support of Iowa State's astronomical observatory, including an interferometer system for measurement of stellar diameters. Dr. Swift is a member of Eta Kappa Nu, Sigma Xi and the IEEE. O April 20,1987 Page 62

l i pd SAM C. YEAGER President, Automated Controls Technology Division B.S., Accounting, Harding College; CPA, State of Georgia Mr. Yeager joined IRT in 1984 through the acquisition of Ridge, Inc. where he served as President and Chief Executive Officer, as well as being one of the major shareholders. Mr. Yeager's career with Ridge began in 1976 after two years with Arthur Andersen & Co. in their Atlanta office. His early jobs with Ridge included office management,' project mangement, and sales and marketing. Mr. Yeager was responsible for building a nationwide sales force at Ridge which allowed the company to grow from $3 million in 1977 to $10 million in 1982. In 1982 he was promoted to Vice President, Sales and Marketing. Recognizing the need for Ridge to have proprietary products, he was responsible for Ridge's decision to design, develop, and manufacture an Industrial Microfocus X-Ray system. This equipment distinguished Ridge from the other industrial x-ray companies, and Ridge soon became the largest industrial x-ray supplier in the United States. After the acquisition of Ridge by IRT, Mr. Yeager retained P/L responsibility for Ridge. In September 1986, he moved to San Diego to take over the responsibility for all U custom inspection and process control activities at IRT, including Ridge. This division is now called the Automated Controls Technology Division and includes approximately 150 employees, with sales revenues projected for FY 87 at $22,000,000. l l l l O lL April 20,1987 Page 63 l

CHARLES A. PRESKITT B.S., M.S., and Ph.D., Physics, University of Rochester In his present position of Chief Scientist, Nuclear Systems Division, Dr. Preskitt makes technical contributions to many IRT programs and is responsible for the technical review of a wide range of IRT scientific and engineering expertise that is applied to the problems of government and industry. In his own specialty of measurement technology he has worked extensively in the field of experimental technique development and statistical data interpretation, and conducts tutorial efforts to keep the IRT technical staff familiar with advanced techniques in this area. Dr. Preskitt has been active in the American Nuclear Society for ten years and was Chairman of the Reactor Physics Division of the American Nuclear Society for the 1976-77 year. As a member of the staff since 1968, Dr. Preskitt has been the program manager and the pincipal technical contributor on a number of large study and development programs. Included are programs to develop advanced materials measurement equip-ment, security systems for the nuclear industry, neutron radiography and radiation gauging systems for _ nondestructive testing applications. He managed a major study and experimental effort on pyrometallurgical reprocessing of nuclear fuels, and a tech-nology assessment program related to instrument requirements for advanced nuclear fuel reprocessing facilities. l i O April 20,1987 Page 64

c-K. L. CROSBIE, P.E. ' B.ChE., Chemical Engineering,' Pratt Institute; M.N.E., N=-laar Ergir uing,. University of Virginia Mr. Crosbie joined IRT in 1963. He was Radiation Safety Officer for IRT from ~ April 1973 to May 1983'and was responsible for all aspects of the radiation safety programs, radioactive materials and special nuclear materials licensing, criticality safety,' and special nuclear materials safeguards and accountability. In addition to these duties, Mr. Crosbie is a Principal Engineer in the Nuclear Systems Division and Principal Investigator for programs involving the design, construction, operational testing and installation of radiation gauging devices and radiation detection systems. Other responsibilities involved the design, construction,' operational testing and installa-tion of Californium-252 based subcritical multiplier systems and physics and engineer-ing investigation on programs related to in-depth heating studies in concrete using microwave and fission heat sources, nuclear instrumentation and radiation shielding. Prior to this he was engaged in fast-neutron physics and fast-neutron detection - systems employing time-of-flight techniques. He was responsible for the design, construction, licensing, and operation of Accelerator Pulsed Fast Assemblies (APFA-I, II, and Ill). As supervisor of these facilities, he has provided static and pulsed service f ' irradiations for' studies that ranged from effects of simulated nuclear detonations to. l foil activation for fast neutron dosimetry. Mr. Crosbie attended the University of Virginia under an AEC fellowship program i and conducted research and development work on a nuclear-powered plasma thermo-1 l couple. From 1961 to 1963, he was with the Sandia Corporation as a reactor supervisor at the pulsed-reactor facility.. His main responsibilities included the ' supervision of pulsed-reactor operations, with primary concern for the safety of the reactor and l personnel conducting experiments. He was also engaged in research on the penetration - and activation of shielding materials by neutrons from the reactor pulse. l r l i l O April 20,1987 Page 65 w.-. - ,,4--w,.ze.v.%-y-%,.- re__-.m,,,,e m- - r .m.m -y ---,--m_

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PAUL R. MASCHKA Mr. Maschka was appointed Radiation Safety Officer for IRT Corporation in May 1983. Prior to that date he was the Senior Health Physicist for IRT since joining the company in May 1973. As Radiation Safety Officer he is responsible for licensing radioactive materials and special nuclear materials; registration of cabinet radiography systems; compliance with the various State and Federal Regulations; and all aspects of the radiation safety program. As Nuclear Materials Manager, he is responsible for special nuclear materials safeguards and accountability and criticality safety. He assists in the planning for all radiation operations, and monitors the projects for compliance, and conducts the Radiological Safety Course for IRT and customer personnel. Prior to joining IRT Corporation, Mr. Maschka was a member of the Health Physics Department of G.A. Technologies, Inc. Since 1963 he has had extensive experience as Health Physics Surveyor at various G.A. facilities, including the Hot Cell, the TRIGA Reactor Facility, Radiation Research Laboratories, and the Fuel Manufac-turing Facilities. In September 1970 he was appointed Senior Health Physicist for the Linear Accelerator Facility, which is now owned by IRT Corporation. Mr. Maschka completed two years of higher education at Creighton University in Omaha, Nebraska. In 1955 he enlisted in the U. S. Air Force and served four years as an electronics technician. In 1959 he was chosen to attend the U. S. Army's Nuclear Power Training Course at Ft. Belvoir, Virginia. He graduated from the 48-week course in March 1961, specializing in Health Physics and Chemistry. Mr. Maschka was a member of the Startup Crew for the U. S. Navy's PM3A nuclear power plant in McMurdo, Antarctica. At the PM3A he worked as Health Physicist, Reactor Operator, and eventually as Shift Supervisor. He received training in nuclear materials management and NMMSS reporting from a seminar conducted by the doe in August 1980 and a workshop conducted for the NRC by Martin-Marietta Energy Systems in July 1985. O' April 20, f.987 Page 66

c- \\ 'O STANLEY 3. FRIESENHAHN B.S., Physics, St. Mary's University; M.S., Physics, University of Notre Dame Mr. Friesenhahn was associated with the company's neutron capture cross-section program from 1961 to 1970. He assisted in the design and construction of the large liquid scintillators and associated apparatus used in the capture measurements. He also assisted in the determination of resonance parameters for sodium, zirconium, erbium, thorium, rhenium, and gadolinium. He performed measurements of the low-energy capture cross section of the isotopes of tungsten, zirconium, and rhenium. In addition he has made measurements in the kilovolt region of the average capture cross sections 238 of gold, rhenium, gadolinium, molybdenum, tantalum, and U. More recently, 10 'Mr. Friesenhahn assisted in a program to compare the absorption cross sections of B and H with the H-scattering cross section from 10 to 400 kev. He was also in charge 10 (n, 0)7Li* cross sections with a B of a series of measurements to determine the precision of 1 to 3% from i kev to 1 MeV, using the hydrogen-scattering cross sections as a flux standard. In recent years he has been engaged in a program to measure 239 5 fission product decay heat for.235U and Pu from 1 to 10 s. He has also measured the beta and gamma spectra as a function of cooling time for those same isotopes. Additionally, Mr. Friesenhahn serves as the chairman of the IRT radiological safety committee. Previously, Mr. Friesenhahn was concerned with lon chamber measurements of subthreshold fission cross sections of the transuranic elements, using the crystal spectrometer facility of the Hanford Laboratories. He also assisted with the start up of the double-crystal neutron spectrometer at the Hanford K-East reactor. He is currently engaged in a program to measure the radiation dose to a reactor containment vessel using proton recoil spectrometry. Recent activities involve development of a low-cost, large-area neutron detector (patent applied for) and the demonstration of this new detector in a neutron portal radiation monitor. He has developed programs for the INTEL 4004 and 8085 micropro-cessors for portal radiation monitor and whole-body counting applications. In addition, he has augmented the capabilities of the High-Level Neutron Coincidence Counter system via improvements to the Motorola 6800 microprocessor program. In addition, he has developed a commercial version of the Brookhaven Survey Assay Meter for nuclear fuel enrichment measurement applications. April 20,1987 Page 67

JOHN W. HARRITY B.S. and M.S., Physics, San Diego State University Mr. Harrity, who joined the staff of IRT in 1958, has served as head of the Test and Services Group since 1973. His responsibilities include management of all ionization and displacement effects tests on semiconductor piece parts, data processing of the test results, and data interpretation. He is presently conducting several programs on latchup screening of junction-isolated integrated circuits and lot qualifica-tion testing of all types of semiconductor devices. Mr. Harrity is IRT's primary consultant on radiation tests conducted on subsystems and systems both at IRT and at Government facilities throughout the country. He is presently a member of the ASTM Radiation Effects Standards Committee. Mr. Harrity has served as Principal Investigator for research programs on displacement-damage radiation effects on semiconductor materials and studies of the transient effects of radiation on germanium, silicon, and a variety of dielectric materials. For several years he was engaged in the testing of semiconductor devices and electronic circuits, with particular attention to the effects of bursts of ionizing radiation on the performance of silicon monolithic integrated circuits. He was Principal Investigator on projects investigating short-term annealing effects in silicon exposed to a neutron burst and on projects studying the effects of ionizing radiation on dielectric materials. He directed a project in the New Point underground nuclear test at the AEC Nevada Test Site. From 1956 to 1957, Mr. Harrity worked at the Visibility Laboratory of the Scripps Institute of Oceanography performing mathematical analyses of optical systems and photoelectric simulators of human scanning systems. s O April 20,1987 Page 68

T d DOUGLAS E. WILLIS 4 B.S.E.E., Mm==achusetts Institute of Technology Since_ Joining the staff in 1960, Mr. Willis has been actively engaged in all phases of radiation-effects work. He conducted some of the first vulnerability studies on military systems, and his early experimental work on the effects of radiation on semi-conductor devices and on coaxial cables provided a basis for the studies presently being conducted. He has successfully conducted a number of experiments in radiation effects at the Nevada Test Site. These experiments involved tests from the piece-part level up through the system level. As a senior member of the Test and Services Group, he directed ionization tests of all electronic piece parts at the linear accelerator facility. In this role he was responsible for major developments in highly automated techniques for device testing in a radiation environment. Mr. Willis was promoted to manager of the linear accelerator facility in 1975. In this capacity he was responsible for the operation, maintenance and modification to the accelerator. In addition, he was also principal investigator on a number of projects associated with the use of the accelerator. In 1986 he returned full time to the Electronics Test Group. Mr. Willis spent four years at General Dynamics / Fort Worth as a design engineer for the nuclear group. His work included the design of electrical controls for reactors and of the electrical equipment for radiation effects studies. April 20,1987 Page 69

1 3ACK C. YOUNG, P.E. B.A., Physics and Mathematics, North Texas State University M.A., Physics, North Texas State University Since joining IRT in 1958, Mr. Young has been engaged in many low-energy neutron physics problems, including work in studies of gamma-ray emission after uranium photofission, the measurement of the Fermi age to indium resonance in BeO, as l well as numerous experiments in integral neutron thermalization. He was Physicist-in-Charge of the Critical Assembly which was used to investigate neutron spectra in various nonmultiplying and subcritical assembles. In 1967, he worked on the design and construction of a thermionic critical facility. In 1968, he was involved in making spectrum measurements in fast subcritical assemblies, with relatively long dieaway times, to develop experimental and analytical techniques for time-of-flight measure-ments of fast neutron spectra. He was Program Manager and Principal Investigator on this program from 1969 to 1972. Since 1973, Mr. Young has been inve,1ved in the management of several integral measurements to test neutron scattering cross sections and gamma-ray production in various shielding materials. Mr. Young was prompted to Manager of the linear accelerator facility in 1986. In this capacity he is responsible for the operation, maintenance and modification to the accelerator. Mr. Young has served as Program Manager and Principal Investigator on several DoD programs, including the development of a fully automated ordnance inspection system for the demilitarization of practica bombs via dual Compton-scattering methods, the fully automated precision measurement of components for SSPO via the Radiation Photometry Gauge (RPG-100), of which two systems were delivered, and several commercial systems, including the Active Fuel Rod Scanner (AFRS-110) which automatically measures the enrichment uniformity, total fissile content, rod and plenum length, fuel stack length, and other important parameters. Delayed gamma sensing of fissions caused by low-level neutron irradiation was utilized in this system. In all of these systems Mr. Young was responsible for the overall system performance, cost and schedule, as well as system integration and day-to-day project management. He has been very successful at taking R&D or conceptual designs to full-throughput, production-line inspection machines. O April 20,1987 Page 70

c i \\ V 9.4 FACILITIES IRT Corporation has two facilities for conducting R&D and product development programs which utilize special nuclear materials. These facilities are described below. E 9.4.1 3030 Callan Road The headquarters and main engineering and scientific building is located at 3030 Callan Road in the Torrey Pines Science Park. The building is a three-story, m 86,500 square foot building with the main entrance on the second floor and truck access / loading on the first floor. The laboratories and shops are located on the first floor. No radioactive materials other than small check sources used for instructional purposes shall be allowed on the upper floors. Radioactive materials may be used throughout the entire First Floor (see Figure 5). SNM will be used in Rooms 101,107,111,113,114,155, and 117. SNM will be sto' ed in the Vault Room 110, and byproduct material will be stored and used in r Room 111. All use of radioacive material and SNM will be conducted in such a way that the second floor shall remain an unrestricted area with personnel exposures limited to ) the levels specified in 10 CFR 10.105. The floors, other than the support structure, are made of 3/4-inch plywood with an overlay of 1-1/2 inches of normal concrete. Since this thickness of wood and concrete offers very little in the way of shielding, all experiments and equipment will be designed in such a way that there will be a maximum amount of shielding on the top in order to keep second floor personnel exposures within the limits of 10 CFR 20.105 and ALARA. The byproduct material storage room has a 10 inch diameter by 10 foot deep pipe sunk into the floor. This will be a storage well for highly radioactive sources. The sources will be stored in individual pipes that are fastened to a plate on top of the well. A movable shield cap of concrete and lead will be positioned over the well. Other sources will be stored in shipping / storage casks and a lead cave of adequate shielding to protect personnel on the second floor. Storage of SNM will be as described previously in Section 6.8. In order to prevent exposure to the office personnel on the upper floors, all movements and use of radioactive materials will be done in compliance with 10 CFR 20.105. If it is calculated that a certain operation could result in an individual on an upper floor receiving a total dose of 2 millirems in any one hour, the following procedure will be followed. April 20,1987 Page 71

It will be determined if the area on the second floor above the radiation work area can be cleared of personnel during normal working hours; if not, the work will be done af ter normal working hours when fewer people e.re present. In either case, during or after working hours, the procedure will be as follows. 1. Clear the area of personnel. 2. Warn the personnel in the adjacent areas, or at night, warn the maintenance personnel. 3. Put up temporary barriers and warning signs around the area on the upper floor. 4. Post an observer, if necessary, to warn people to stay out of the area. 5. At the end of the operatien, remove the ropes and the signs and notify personnel that it is safe to return. If the source that has been installed in a device or if a radiation-producing machine is to be operated, the area on the upper ficor will be surveyed with the appropriate portable radiation survey meters in order to determine the dose rates in that area. If the survey indicates that the limits imposed by 10 CFR 20.105 might be exceeded, additional shielding material will be added to the device or machine to reduce the dose rates to acceptable levels. The radiation dose rates will be verified by additional surveys. Only then will step 5 above be taken. Environmental and location badges shall be placed in normally occupied areas on the Second Floor above those areas where radioactive materials are used or stored on the First Floor. The badges are placed in the following areas (refer to Figure 6): Room 200 - Accounting Room 221 - Lobby Room 234 - Publications and other areas as needed. We have had.three years of experience and film badge data. The highest yearly exposure at any one location was 140 mrem. Since personnel only work about 2000 hours a year (about one-fourth of the year), the maximum calculated personnel exposure is 35 mrem for the year. O April 20,1987 Page 72

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9.4.2 8221 Arjons Road This manufacturing and warehouse facility is located in an industrial area North of the Miramar Naval Air Station at the North end of a long warehouse complex. Other companies occupy the rest of the building to the South of this facility. The building is made of tilt-up concrete slabs 20 feet high. In part of the facility there is a drop ceiling eight feet high, but in the CXI Assembly and SNM storage areas the ceiling is the fully 20 feet high. The facility measures 180 feet by 160 feet. Only SNM contained in sealed sources will be used at this facility. No materials with smearable ~ contamination in excess of 66 dpm/100 cmsg Beta / Gamma or 12 dpm/100 cmsq Alpha will be allowed to be used at this facility. The facility is used to manufacture Portal Radiation Monitors (PRMs), Portal Beta Monitors (PBMs), Automated Contamination Monitors (ACMs), and Component X-Ray Inspection Devices (CXIs). This facility is also being used for R&D to design, build, and test new versions of the above devices and other devices, such as sincillation detector systems, SNM monitoring and analyzing equipment, and neutron radiography. Other than the Pu/U rods, the SNM shall be stored in the scintillation room inside a locked cabinet. The Pu/U mixed oxide fuel rods shall be stored in the Storage Vault Room as described previously in Section 6.8.3. 9.5 EFFLUENT CONTROL Effluent is controlled within the regulations as required by 10 CFR 20 with the additional requirements as previously stated. 9.5.1 Air Programs involving the use of radioactive material in a form that may become airborne are connected to or carried out within areas that have a filtered ventilation system. 9.5.2 Liquid Liquid effluent is controlled by storage until determined safe for disposal or is safe within applicable regulations for release to a sanitary sewage system. April 20,1987 Page 76

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Monitoring of effluents is accomplished by using portable air samplers and continuous air monitors and performing analysis for long-lived alpha and beta radioac-tivity and by gross-alpha, gross-beta and gamma spectral analysis of liquid samples. 9.6 INSTRUMENTATION The Radiation Safety Office has instrumentation available to perform the following functions in the detection and measurement of radiation. 9.6.1 Personnel Monitoring Devices IRT uses photographic dosimetry for x-ray and gamma radiation in the range 5 kev to 100 MeV covering dosages from 20 mrem to 500 rem for high energies, and dosages from 10 mrem to 60 rem for low energies; photographic dosimetry of beta radiation from 400 kev to 50 MeV covering dosages from 40 mrem to 1000 rem; thermal and fast neutron photographic dosimetry in the ranges 0 to 3.5 eV covering dosages from 10 mrem to 300 rem, and 500 kev to over 100 MeV covering dosages from 15 mrem to 10 rem. Pocket lonization-chamber dosimeters wh.ch are the self-reading type that detect x-ray and gamma radiation from a dose of 1 mrem to 500 rem, and self-reading fast neutron and thermal neutron dosimeters that measure dosages from 1 mrem to 200 mrem are also used. LiF and CaF TLD's are available to the Radiation Safety Office on loan from other IRT operating departments. 9.6.2 Radiation Monitoring and Survey Instruments A list of instrumentation, detection limits, and frequency of calibration is provided in the table on the following page. 9.6.3 Radioactive Material Assay The Radiation Safety Office has the following Radioactive Material Assay instruments. 1. An Eberline Model BC-4 Counter. This instrument is a complete system consisting of a two-inch detector, high-voltage power supply, pulse amplifier, timer, and six-decade scaler. All circuits are solid state with extensive use of integrated circuits to enhance reliability. The detector is a pancake-type Geiger tube with a 1.75-inch diameter window which is 1.4 to 2.0 mg/cm2 April 20,1987 Fage 78

thick. With the addition of a 0.002-inch mylar window the total window density is approximately 7 mg/cm2 This mylar window is removable for decontamination or for counting low energy beta emitters such as I"C. the detector is shielded on the top and side with 7/8 inch of lead. The detector . background is generally 30 to 33 cpm. The efficiency for two pi geometry is . generally 70 percent and for four pi the efficiency is 35 percent for 90Sr 90, y 2. An Ebe.rline Model SAC-4 Scintillation Alpha Counter. This instrument is a complete system consisting of a two-inch detector, high-voltage power supply, charge-sensitive input amplifier, timer, and six-decade readout. All circuits are solid state, except the detector, with extensive use of integrated circuits to enhance reliability. The detector is scintillation phosphor made of ZnS(Ag) powder plated on a plastic light pipe. The photomultiplier tube is a ten-stage, Sil-response, end window tube, two inches in diameter. The background is generally 0.6 to 0.9 cpm. The efficiency for two pi geometry is generally 80 percent for a one-inch diameter Pu239 source, and for four pl O geometry the efficiency is 40 percent for the above source. %.) 4 April 20,1987 Page 79

Minimum Detection Detection Calibration Type Limits Limits Frequency G.M. Survey Meter Ludlum 14A (5) 0-50,000 cpm 100 cpm Annually Ludlum 12 (1) 0-500,000 cpm or 2 x B.G. l Technical Associates (TBM-1) 0-50,000 cpm { Alpha Survey Meter Eberline PAC-ISA (2) 0-2,000,000 cpm 100 cpm Quarterly Ludlum Model 12 (1) 0-500,000 cpm 150 cpm or 2 x B.G. Neutron Survey Meter i l Snoopy, NP-1 and 0-2,000 mrem /hr 0.2 mrem /hr Quarterly NP-2 (3) Ionization Type Survey Meter Eberline RO-2 (1) 0-5,000 mr/hr 0.1 mr/hr Quarterly Victoreen 440 RF (1) 0-300 mr/hr 0.1 mr/hr Quarterly f Eberline RO-2A 0-50,000 mr/hr 0.2 mr/hr Quarterly Victoreen 592B (1) 0-1,000 mr/hr 0.5 mr/hr Quarterly l Victoreen 471 RF (1) 0-300 R/hr 0.02 mr/hr Quarterly Criticality Monitor GM Detectors 0-20 mr/hr 2 mr/hr Quarterly Eberline RM-12A (3) I 3. A Technical Associates windowless gas flow proportional comter for assaying Tritium (3 ), Alpha, and Beta radiations. The system consists of a power H supply, an automatic scaler, a timer, a scanning single-channel analyzer, and the windowless gas flow counter. The detector is a two-inch diameter, two-pi hemisphere with a precision collector wire. The two-pi efficiency is normally 51 percent for alphas and 60 to 70 percent for betas. The Radiation Safety Office also has free access to the following equipment used I i by other groups at IRT. l 1. For sample counting: a Tennelec LB1000 Low Beta counter and an ORTEC 1 576 Alpha Spectrometer. 2. For sample assay and isotope identification: there are available a number of I Na! detectors and Ge(Li) detectors used with TMC or Nuclear Data 660 multichannel analyzers. April 20,1987 Page 80

r 9.6.4 Air Samples The' Radiation Safety Office has portable particulate air samplers that can use j various types of filter papers. Each of the air samplers includes a gas flow meter for. the determination of the amount of air that flows through the filter. The filter papers are analyzed for alpha, beta, and gamma radiation. 1 . i t I i l-i i lO i l April 20,1987 Page 81 j l

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(V \\ \\ \\ - 10. MANUFACTURING AND QUALITY ASSURANCE PROCEDURES IRT Corporation manufactures devices such as fuel rod and fuel pellet scanners, Special Nuclear Material active and passive assay systems, neutron and gamma gauging systems, narcotics detectors, letter bomb detectors, and subcritical systems. Distribu-tion is to specific licensees or _ DOE /NRC-exempt contractors. Except for. sources which are securely attached to the devices and meet the requirements set forth in 49 CFR, Section 173.391(b), the isotopic sources are shipped separate from the device, in accordance with current DOT regulations. These devices are manufactured at IRT facilities under the provisions of the Quality Assurance Program and performance tested for acceptance and to verify customers performance specification. Acceptance testing is generally performed in the presence of customer quality assurance personnel. Prior. to shipment, the sources are tested for contamination and leakage. Results are documented on the transfer and shipping documentation to the customer. Upon receipt of the source, the responsibility falls upon the specific !!censee. t i I i t-i O April 20,1987 e i Page 83 i

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11. WASTE DISPOSAL Waste disposal service is on an as-needed basis with Thomas Gray and Associates of Orange, California. Waste processing is not done by IRT Corporation; accumulated I wastes, liquid and solid, are packaged and labeled in accordance with applicable DOT regulations and transferred to the Broker for ultimate processing and disposal.

11.1 SOLID WASTE DISPOSAL Very little solid waste is generated by the operations in which IRT Corporation is presently involved. Over the past several years, most radioactive waste has come from cleanup of high-energy accelerator operations that have long since been terminated. ( 11.1.1 Work Area Waste Receptacles Approved waste receptacles are metal trash cans equipped with a self-closing tid or a metal drum with a full metal cover. A plastic bag is placed inside each receptacle with the open end of the bag pulled down the outside of the can leaving the bottom of the bag sitting on the bottom of the receptacle. (This helps prevent contamination of the outside of the can.) An inventory sheet is attached to the can. When radioactive waste is placed in the receptacle, the isotope and amount (if known) are noted on the inventory sheet. SNM waste is segregated from byproduct material waste and collected in a separate receptacle. If a number of different isotopes are used in the area, two waste receptacles are used, one for radioactive isotopes with half-lives less than 30 days, and another for Isotopes with half-lives greater than 30 days. 11.1.2 Collection of Waste The Waste Disposal Personnel are notified when a waste receptacle is full or nearly full. The waste disposal personnel remove the bag by gathering up the open end of the bag and twisting it closed just above the mass of waste; the bag is taped shut, and then pulled out of the receptacle. April 20,1987 Page 85

11.1.3 Monitoring The bag is monitored with the appropriate radiation survey meter. Radiation readings are taken at the surface of the bag, at one foot, and at one meter from the bag, with the readings being noted on the inventory sheet. The radiation readings are used to calculate the amount of radioactive material in the bag in case the inventory sheet has not been properly updated. The bag is marked with radiation warning labels and is taken to the waste collection area. 11.1.4 Waste Collection and Storage Area The normal waste collection and storage area is the Radioactive Material Storage l Room at the Callan Road facility. Only DOT 17E/17H drums are authorized for I shipment of Radioactive Waste. Both empty and filled drums are kept under some protective cover to prevent deterioration. Waste containing short-lived isotopes are stored until the radioactive isotopes have decayed to background levels as measured on a pancake G.M. detector, and then the waste is disposed of in the ordinary trash. 11.1.5 Packaging of Shipping Drums A plastic 55-gallon drum liner, at least four mil thick, is placed in the waste drum. The individual bags of waste are put into the drum and packed down loosely taking care not to puncture individual bags or pack so tightly that entrapped air is forced out of the bags. A note is made on the inventory sheet indicating the amount and identity of radioactive materials placed in each drum. The appropriate radioactive warning labels are applied to each drum. 11.1.6 Monitoring Shipping Drums When a drum is filled, the open end of the plastic bag is drawn up and twisted shut just above the mass of waste and secured with a twist tie or tape. The lid is placed on the drum and secured by bolting the locking ring tightly in place. A portable radiation survey meter is us6d to scan completely around the sides of each drum from top to bottom, and, the top and bottom. The highest radiation readings at surface, at one foot, and at one meter are noted on the inventory sheet. At least two wipes are taken on each drum and counted according to standard wipe counting procedures. If a drum is contaminated on the outside it is immediately decontaminated to acceptable April 20,1987 Page 86

M levels. The drum is leak checked by laying it on its side for 20 minutes and noting if liquid comes out around the locking ring. All leaking drums are opened and repackaged. The proper radiation warning labels are placed on the drum and the inventory sheet is taped to the top of the drum. 11.1.7 Disposal of Solid Waste When six drums of waste have accumulated, the commercial hauler is contacted and arrangements are made for removal of the waste. Each drum is classified in accordance with 10 CFR 61.55 and labeled with the appropriate DOT shipping labels. A Radioactive Materials Shipping Record is made out, along with any special papers the hauler may require. A record of all radioactive materials shipped and copies of all papers are stored in the Radioactive Waste Disposal file in accordance with 10 CFR 20. 11.2 LIQUID WASTE DISPOSAL Most liquid wastes have been generated by decoritamination and cleanup of equipment and the accelerator facility. Some low-level concentrations of liquid waste V has been generated by radiochemical operations, and in almost all cases this liquid met the requirements of 10 CFR 20.303 for disposal into the sanitary sewer system. 11.2.1 Work Area Liquid Waste Receptacles Storage containers for collecting liquid waste are leak-proof, corrosion-resistant plastic bottles with screw caps ranging from one to five gallons in capacity. An inventory sheet is maintained for each bottle. SNM, byproduct material, short half-life and long half-life materials are collected in separate bottles. 11.2.2 Monitoring and Collection of Liquid Waste The waste disposal personnel are notified when a bottle is full. They cap the bottle tightly, agitate it vigorously, and extract a one-liter sample. The water sample is assayed by the ' Health Physics Department using standard laboratory procedures; i.e., the water sample is evaporated to dryness and the residue is counted for radioactivity. If it is determined that the liquid meets the requirements of 10 CFR 20.303, it is disposed of by release into the sanitary sewer system. If the liquid j does not meet the requirements of 10 CFR 20.303, it is stored and disposed of by the current U.S. DOT and NRC regulations. April 20,1987 Page 87

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12. CERTIFICATE The applicant and any official executing this certificate on behalf of the applicant-named in Item 1, certify that all information contained herein, including any supple-ments attached hereto, is true and correct.

IRT Corporation Applicant Name in Item 1 Date: May 22, 1987 By: - Sam Yeager Vice President ?- Title of Certifying Offipt/ i O April 20,1987 Page 89 l

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O APPENDIX I SUPPLEMENTAL INFORMATION CONCERNING PROPOSED USES O l April 20,1987 Page 91 ~

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1. RESEARCH AND DEVELOPMENT PROGRAMS These programs sometimes involve the use of unsealed and/or liquid sources.

These programs are conducted under the following general safety considerations. 1.1 SAFETY The ~ IRT Radiation Safety Committee, which reviews all in-house programs utilizing radioactive materials, has estabilshed a set of general safety rules for handling uncontained radioisotopes. 1. If at all feasible, operations should be carried out within the confines of a hood or glove box, even if the quantities are sufficiently small to warrant open bench-top operations. AU 2. All operations involving finely divided particles of pyrophoric radionuclides must be carried out under an inert atmosphere within a gloved box. Storage of these materials must be in fireproof containers. 3. Keep work area free of all unnecessary equipment and cover work area with protective absorbent paper. If possible, carry out operations within con-tainers or catch trays and keep tools, equipment, etc., in localized area within suitable containers. 4. Plan operations to minimize handling and transfer of materials and amount of material used. 5. Keep waste generation to a minimum and maintain waste containers in immediate area; keep SNM and byproduct wastes segregated. Contact Health Physicist.for specific instructions for disposition. 6. Use protective clothing as necessary and monitor self prior to leaving work area. l 7. Do not handle loose material directly; use tongs, tweezers, pipettes, etc. Do g use mouth technique for pipetting operations. April 20,1987 Page 93

8. Make routine contamination surveys daily until operational techniques are perfected, and then, if warranted, on a weekly basis. Send wipe samples to Health Physics for analysis. 9. Use portable air sampler in vicinity of operations whenever operations are in progress. Send samples to Health Physics on daily basis for analysis. 10. Store materials when not in immediate use in hood and return to Health Physicist for storage when operations are completed. Store materials in closed metal cans and, if liquids, include enough absorbent material in container to fully absorb all material. 11. In case of spills or any unusual events, contact Health Physicist or Radiation Safety Officer for assistance. Other more specific requirements may be imposed by the Radiation Safety Committee contingent upon the particular program. As a general rule, the Health Physicist or Radiation Safety Officer will be present for the initial operations of any nonroutine program. In determining the amounts of materials which may safely be used outside a hood or glove box or within the confines of a hood, the Radiation Safety Committee follows the guideline "Workplaces for Unsealed Radionuclides," authored by Mr. D. Pickler of the State of California Radiologic Health Unit. The guideline is included in Appendix III. I O April 20,1987 Page 94

O APPENDIX II RADIATION WORK AUTHORIZATION FORM O O April 20,1987 Page 95

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e IRT Corporation RADIATION WORK AUTHORIZATION NO. s ) (,/ 1. ( ) New Request Date Submitted ( ) Renewal of RWA No. Date Required 2. Principal Investigator: Material Custodian Alternate 3. Work location 4. Description of program (include methods, sketch of the apparatus and setup showing the shielding provided). Attach supplements referenced to this request,if necessary. <3 ) a 5. Quantities of Radioactive Material involved in this operation: Isotope Amount Material Description (Physical Form) 6. If Airborne Radioactivity is created indicate major Isotopes, quantity and physical form. Isotope Amount Gaseous or Particulate NONE 7. Will a contamination control boundary be needed: ( ) Yes ( ) No ] If yes, describe. Attach supplements referenced to this request,if necessary. April 20,1987 Page 97 3

8. Is Special Nucl:ar Mat: rial to be used under this RWA? ( ) Yes ( ) N3 If Yes, specify use and storage locations. Total throughput of special nuclear material during one-year period: ( ) grams U-235; ( ) grams plutonium; ( ) grams U-233 g Form in which material will be returned to accountability 9. Upon receipt of the radioactive material to be utilized as described herein, (I) (We) shall be re-sponsible to maintain the exposure of any individual to the radiation therefrom to as low as prac-ticable limits. (1) (We) have read, are familiar with, and will comply with Title 10, Part 20 Code of Federal Regulations (Standards for Protection Against Radiation), and Title 17, Chapter 3, Sub-chapter 4, Group 3, California Administrative Code (Standard for Protection Against Radiation), IRT Radiological Safety Guide, (if applicable) the Linac Radiological Safety Regulations, and comments under Section 11. (I) (We) have successfully passed the Radiological Safety Test: Principal Investigator Material Custodian Alternate Material Custodian ADDITIONAL AUTHORIZED PERSONNEL 10. Take to H.P. for discussion and approval. 11. Radiation Safety Committee Comments: 12. Approvals: Cognizant Manager: Date Health Physicist: Date Radiation Safety Officer: Date Chairman, Criticality Safety Committee: Date Chairman, Radiation Safety Committee: Date This Approval Expires on: Frequency of auditing 13. Have authorized personnel read and sign above. Return to Radiation Safety Officer. April 20,1987 { Page 98 2

O APPENDIX 111 WORKPLACES FOR UNSEALED RADIONUCLIDES O O Aprl! 20,1987 Page 99

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m l '\\. V APPENDIX III WORKPLACES FOR UNSEALED RADIONUCLIDES by D. A. Pickler A search was made for existing guidelines as the maximum amount of unsealed radioactive material which can be safely used: (1)outside hoods or glove boxes; and (2) in hoods. If the guidelines examined (Refs. B, C, D, and H), only References E and H are definite in this aspect. The latter part of this paper proposes a new guideline which, it is believed, has advantages over those of ' he referenced publications. t Reference A discusses the toxicity of radionuclides which may become incorpor-ated in the human body. Of special interest is the conclusion that in working with radionuclides, inhalation, rather than ingestion, is the most significant mode of entry in the body because ingestion is usually more readily controlled or avolded by taking simple precautions. Accordingly, in that document radionuc!! des are classified as to toxicity based on two factors--the MPS (annual inhalation maximum permissible intake) in microcuries, and the MPI in micrograms. The MPI in microcuries is the only factor used for radionuclides of higher specific activity. With radionuclides of lower specific activity, both factors are used; the MPI in micrograms being used to " downgrade" the toxicity because of the smaller probability of breathing significantly harmful amounts of radioactive material with lower specific activity. All radionuclides with an MPI of over 10 milligrams are classified as " low" toxicity. All radionuclides with an MPI of between 0.1 and 10 milligrams would be classified as " upper medium" toxicity. Thus, for example, strontium-90 is classified as "high" toxicity and natural uranium and natural thorium are classified as " low" toxicity, although the MPI in microcuries is greater for strontium-90 than for natural uranium or natural thorium. References B and C are very similar to each other. Apparently, Reference C was based directly or indirectly on Reference B. Both contain the following table. I 10 April 20,1987 Page 101 l

O Type of Laboratory or Working Place Required Type C Relative Minimum Good Type B Type A Radiotoxicity Significant Chemical Radioisotope High Level ofisotopes Quantity Laboratory Laboratory Laboratory Very high 0.1 pCl 10 pCl or less 10 pCi-10 mCl 10 mci or more High 1.0 pCi 100 pCl or less 100 C1 - 100 mCl 100 mCl or more Moderate 10 pCi 1 mCl or less 1 mCl - 1 Ci 1 Ci or more Slight 100 Cl 10 mCl or less 10 mCl-10 Cl 10 Cl or more References B and C also contain a statement, as follows, on modifying factors to be used with the above table. Modifying Procedure of Operation Factor Storage X 100 Very simple wet operations X 10 Normal chemical operations X 1 Complex wet operations with risk of skills X 0.1 Simple dry operations X 0.1 Dry and dusty operations, and those where X 0.01 Isotopes are evolved as gases Each of the more generally used radionuclides is listed under one of the four radiotoxicity classifications in the first column of the table. The method of classifi-cation is not specified, except in a very general manner. In Reference B, the equipment required for Types A, B, and C laboratories or working places is described only in a very general manner. No statement is made as to whether Types B and C facilities should include hoods or glove boxes, or both. The statement is made that "in general, Type A laboratories will use glove boxes or other completely enclosed systems." Reference C ifescribes in considerable detail guidelines for design of a Type B laboratory, including guidelines for design of hoods and glove boxes. Although References B and C provide guidelines for the amount of material which can be safely used in Types A, B, and C facilities, neither provides guidelines for the amount of O April 20,1987 Page 102

c material which may be safely used in hoods for the amount of material which may be safely used in facilities with neither hoods nor glove boxes. Reference D divides the more commonly used radionuclides into four groups, "very high hazard," "high hazard," " medium hazard," and " low hazard" for the soluble forms. It contains no classification for the insoluble forms. For " normal chemical ' operations" a given amount of radionuclide is a " low level" amount, a " medium level" amount, or a "high level" amount (with borderline areas extending for a factor of 10), depending upon the group to which it belongs, as shown in Table 2 of that document. Factors for use other than " normal chemical operations" are the same as for References B and C on one hand, and Reference D on the other; strontium-89 is listed as "high" in References B and C, but as " medium" in Reference D, while sodium-22 is listed as " moderate"in References B and C, but as "high"in Reference D. Reference D states that for low-level work an ordinary fume' hood, such as is used in chemical laboratories, may be used; it also states that for highly toxic or high-level radioactive material, the velocity through hood openings must be 125 to 200 fpm. No amount of radioactive material is specified as small enough for use outside a hood, or as too large for use in a properly designed and operating hood. Reference E also olvides the more commonly used radionuclides into four groups- "very high," "high," " moderate," and " low." It lists some radionuclides not mentioned in References B and C, and does not include some radionuclides listed in those documents. Uranium-233 is classified as "very high" in References B and C, and as "high"in Reference E. Natural uranium is classified as "high"in References B and C, and as " low" In Reference E. Reference E includes the following equation to obtain a guide to the type of workplace required: H = QTU, where H = hazard guide value Q = quantity of radionuclide (in Cl) T = relative toxicity factor U = use factor. Relative toxicity factor (T) is 100 for radionuclides classified as "very high," 10 for _those classified as "high," i for those classified as " moderate," and 0.1 for those classified as " low." April 20,1987 Page 103

o Type of Use Factor Operation (U) Storage 0.01 Very simple, wet 0.10 Normal 1.00 Simple, dry 10.00 Complex, wet 10.00 Dry and dusty 100.00 The workplace required is a function of the hazard guide value (H) as follows: Hazard Guide Value Workplace (H) Required Less than 100 TypeI 100 - 1000 Type II More than 1000 Type III Type I is described in some detail, and is essentially a good laboratory, but with no hood or glove box required. Type II is also described in some detail, including a minimum requirement that operations be carried out in hoods. Type III is also described in some detall, including a minimum requirement that operations be carried out in gloved boxes. Reference H provides a guideline based on eight toxicity groups, but with no allowance for variations according to procedures and operations. The eight groups are chosen to correspond to the eight orders of magnitude over which the estimated maximum doses per curie range for the valrous nuclides when they are delivered in a single intake by inhalation. The following disadvantages are !!sted for the above described methods of establishing guidelines for workplaces for radionuclides: 1. References B, C, and D do not establish guidelines for operations using hoods and glove boxes. 2. Reference E does not assign a use factor gases and volatile materials. Reference H does not assign any use factors. O April 20,1987 Page 104

y n V 3. Radionuc!! des are classified into groups, with each group assigned a relative toxicity factor of 10 greater than that of the less toxic group adjacent to it, resulting in the following conditions: a. Many radionuclides are not listed in any of the groups. b. The classification does not take into account the chemical form of the radionuclide. For example, each guide either considers soluble material only, or there is no distinction between soluble and insoluble forms, although the MPC in air may be quite different for soluble and insoluble material; for radium-226 the MPC in air for soluble and insoluble material is different by a factor of about 6,000. However, care should be exercised in classifying a highly toxic material as " insoluble." For example, radium sulphate might be thought of as " Insoluble," as its solubility is listed (Ref F) as 2 x 10-8 g/cc at 25 C and 5 x 10~8 g/ccat 45 C. If one assumes that the solubility is 4 x 10~8 g/cc at body temperature, and that the amount dissolved in the gastro-intestinal tract n from a single large ingestion of radium-2261s equal to the amount which ) can be dissolved in one liter of water (a purely arbitrary assumption), by using the data in Reference G lt is seen that under these conditions 1.1 pCi of radium-226 (11 times the maximum permissible body burden) could be expected to be deposited in bone from this one ingestion of " insoluble" material. c. Relative toxicity between the most toxic and the least toxic of the four 3 groups of References B, C, D, and E is assigned a value of 10. However, the MPC in air for hydrogen-3 (soluble)is greater than that for 7 plutonium-239 (soluble) by a factor of about 10. To eliminate the above listed disadvantages, it is suggested that a new guide for workplaces for unsealed radionuclides be set up. This guide would, like those discussed above, be for protection against internal exposure only, and would not consider external radiation. The guide would be based on MPCs for air as stated in Handbook 69, except that the criteria for radioactive material of very low specific activity would be different from that for other radioactive material. Handbook 69 is suggested as a basis .for the guide because it is widely distributed and contains information on all commonly b used radionuclides for both soluble and insoluble forms. The exact formulas would have v April 20,1987 Page 105

to be arbitrary to some degree, as are all such formulas. The criteria and formulas below are suggested as such a guide; recommendations are welcomed for modifications of either the criteria or formulas or both. Of course, exceptions to the guide would be made whenever it is reasonable to do so. I. Based on experience, natural thorium, thorium-232, natural uranium, uranium-235, and uranium-238 need not be handled in a hood or glove box provide the material is in a form which is neither volatile nor apparently contains respirable size particles. II. For radioactive material not meeting all the provisions of I above, guides are based on the following, more or less arbitrary, assumptions: 1. Inhalation is the most significant mode of entry into the body. 7 2. Each person breathes 2 x 10 ml of air per day while working with radioactive material (Ref G). 3. The maximum fractional part of radioactive material present which is inhaled per day by a person is as follows: Outside Hood Procedure of Operation or Glove Box In Hood Operations involving gases or volatile 10"I 3 x 10~3 liquids Dry operations with respirable size 10-2 3 x 10~0 particles Dry operations with apparently no 10~3 3 x 10-5 respirable size particles Complex wet operations with risk of 10-3 3 x 10~5 spills Normal chemical operations 10^ 3 x 10-6 Very simple wet operations 10~5 3 x 10~7 Storage of solids and volatile liquids a a Storage of solids and nonvolatile liquids b b aStorage requirements to be based on leakage rate and on maxi-mum credible accident. bStorage requirements to be based on maximum credible accident. April 20,1987 Page 106

m .im. . dsing the above assumptions, one arrives at the following guides (MPC means MPC in pCl/ml in air for a 40-hour week):. Maximum Amount to be Handled Maximum Amount - Outside Hood to be Handled or Glove Box in Hood Procedure or Operation (gCl) (gCl) Operations involving gases of volatile MPC x 10 MPC x 3 x 10' 8 liquids. 9 10 Dry operations with respirable size 'MPC'x 10 MPC x 3 x 10 particles 10 II Dry operations with apparently no MPC x 10 MPC x 3 x 10 respirable size particles Complex wet operations with risk of ' MPC x 10 MPC x 3 x '10II 10 spills II 12 Normal chemical operations MPC x 10 MPC x 3 x.10 12 13' - Very simple wet operations MPC x 10 MPC x 3 x 10 Storage of gases and volatile liquids a a Storage of solids and nonvolatile liquids b b " Storage requirements to be based on leakage rate and on maximum credible accident. bStorage requirements to be based on maximum credible accident., Ill. When radioactive material is handled outside hoods or glove boxes, { Reference E requirements for " Type I workplaces" (paragraph D.l. of I Reference E) will apply. When radioactive material is handled in hoods, i Reference E requirements for " Type II workplaces" (paragraph D.2. of Reference E) will apply. When radioactive materlal is handled in i facilities designed for containment superior to that of hoods, Refer-l ence E requirements for " Type 111 workplaces" (paragraph D.3. of Refer-ence E) will apply. I O April 20,1987 Page 107

References A. A Basic Toxicity Classification of Radionuclides, IAEA Technical Report Series No.15 (1963). B. Safe Handling of Radioisotopes, IAEA Safety Series No.1 (1958). [ C. Design Guide for a Radioisotope Laboratory (Type B), American Standards Associ-ation, Incorporated, sponsored by American Institute of Chemical Engineers (1964). D. National Bureau of Standards Handbook 92. I E. Workplaces for Radionuclides, Lawrence Radiation Laboratory (Livermore) Health Chemistry Manual, Part I, Procedure 702 (1963). F. Handbook of Chemistry and Physics,45th edition, The Chemical Rubber Co. G. Health Physics, June 1960. H. Brodsky, Industrial Hygiene Journal, May-June 1965, p. 294. O l l l I f I I O ) l April 20,1987 l Page 108

t 6 I i I t P i APPENDIX IV OUTLINE--IRT RADIOLOGICAL 5AFETY COURSE IRT RADIOLOGICAL MANUAL IRT 4171-009 i l } I I I i t I i f r i I i 1 i i i April 20,1987 f Page 109 l I 1 l 1 L ~+++ee --w+- W.. w ww P-*

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f .ry d APPENDIX IV OUTLINE--IRT RADIOLOGICAL SAFETY COURSE 1. INTRODUCTION 2. PURPOSE 3. BASIC NUCl. EAR PHENOMENA 3.1 Atomic Structure 3.2 Radiation-3.3 Particulate Radiation 3.4 Electromagnetic Radiation 4. INTERACTION OF RADIATION WITH MATTER 4.1 Ionization \\ 4.2 Absorption or Capture 4.3 Annihilation 4.4 Scattering 4.5 Fission 5. THS NATURE OF R ADIATION 5.1 Alpha Particles 5.2 Beta Particles 5.3 Positrons 5.4 Neutrons 3.4.1 Elastic Scattering 5.4.2 Inelastic Scattering 5.4.3 Absorption 5.5 Gamma Rays and X-Rays 5.5.1 Photoelectric Effect 5.5.2 The Compton Scattering 5.5.3 Pair Production 5.5.4 Gamma-N Reaction 5.6 Protons 5.7 Summary of Radiation Interaction O April 20,1987 Page 111

5.7.1 Alpha Particles 5.7.2 Beta Particles 5.7.3 Positrons 5.7.4 Neutrons 5.7.5 Gamma and X-Rays 5.7.6 Protons 6. UNITS OF RADIATION MEASUREMENT 6.1 Introduction to Radiation Measurement Units 6.2 Definitions of Radiation Units 6.2.1 Roentgen (R) 6.2.2 Roentgen Equivalent Physical (REP) 6.2.3 Radiation Absorbed Dose (RAD) 6.2.4 Relative Biological Effectiveness (RBE), Quality Factor (QF) 6.2.5 Roentgen Equivalent Man (REM) 6.3 Curie (Cl) 6.4 Radioactive Decay 6.5 Half Life 7. BIOLOGICAL EFFECTS OF RADIATION 7.1 Interaction of Radiation with Biological Systems 7.1.1 Direct Method 7.1.2 Indirect Method 7.2 Major Types of Biological Effects 7.2.1 Genetic Effects 7.2.2 Somatic Effects 7.2.3 Specific Organ Response to Radiation 7.3 Biological Experience of Radiation Damage 7.4 Radiation Exposure Limits 7.5 Background Radiation Exposure 7.6 Effects of Whole-Body Doses Received in a Short Time 7.7 Risk of Biological Effects 8. PRINCIPLES OF RADIATION DETECTION AND MONITORING TECHNIQUES 8.1 fonization Detectors 8.1.1 lon Chambers 8.1.2 Proportional Counters 8.1.3 Geiger Counters 8.1.4 Summary O April 20,1987 Page 112

e 8.2 Scintillation Detectors 8.3 Secondary Particle Detectors 8.4 Photographic Detection Method . 8.5 Thermoluminescent Dosimeters 8.6 Personnel Monitoring Equipment 8.6.1 Film Badges 3.6.2 Pocket Dosimeters 8.6.3 Chirpers 8.7 Survey Instruments 8.7.1 Geiger Counters 8.7.2 Juno Survey Meter 8.7.3 Victoreen 440 RF Survey Meter 8.7.4 Victoreen 592B 8.7.5 The Snoopy 8.7.6 Alpha Survey Meters 8.8 Monitoring Techniques 8.8.1 Detection Operations 8.8.2 Monitoring Operations 9. RADIATION PROTECTION \\ 9.1 External Radiation Protection 9.1.1 Time 9.1.2 Distance 9.1.3 Shielding 9.2 Internal Radiation Protection 9.3 Safety and Handling Generations 10. RULES AND REGULATIONS 10.1 Title 10 CFR 19 10.2 Title 10 CFR 20 10.2.1 Exposure Limits 10.2.2 Personnel Monitoring 10.2.3 Signs 10.2.4 Receipt of Radioactive Material 10.2.5 Incidents 10.3 License Requirements 10.3.1 The Radiation Safety Committee (RSC) 10.;.1 The Criticality Safety Committee (CSC) 10.3.3 The Radiation Safety Officer (RSO) O 10.3.4 The Health Physicist (HP) April 20,1987 Page 113

11. THE LINEAR ELECTRON ACCLELERATOR 11.1 Fundamental Linac Radiological Safety Rules 12. CALIFORNIUM-252 13. RULES FOR SAFETY 14. RADIATION WORK AUTHORIZATION 14.1 Procedures for Initiating an RWA 15. SHIPMENT OF RADIOACTIVE MATERIALS 15.1 Shipments Between IRT's San Diego Facilities 15.2 Shipments to Other Companies 15.3 Shipment by Air 15.4 Shipment of Special Nuclear Materials 16. DEFINITIONS O O April 20,1987 Page 114

r. ~ ..O APPENDIX Y SUPPLEMENTAL INFORMATION FOR THE USE OF UNSEALED SOLID MATERIAL IN NONDISPERSIBLE FORM i i 'O April 20,1987 Page 115 I ..-,--,.---._.m ..._,.,.,,c,

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. j]N ( APPENDIX V l SUPPLEMENTAL INFORMATION FOR THE USE OF UNSEALED SOLID MATERIAL IN NONDISPERSIBLE FORM

1. PROGRAMS INVOLVING SOLID MATERIALS IN NONDISPERSIBLE FORM This supplement describes a specific program that involves loading sintered fuel pellets into rods that will be used to test an automatic fuel rod scanner. The general safeguards, radiation safety, receipt and shipping procedures described herein will be followed for all other programs involving not more than 300 grams of solid materials in nondispersible form.

2. ASSEMBLY OF TEST RODS FOR THE FUEL ROD SCANNER 235 This program involves the use of up to 700 grams of U fuel pellets with enrichments of 2% to 4%, with no more than 300 grams of material being opened and unsealed at any one time, f}

b 2.1 RECEIPT OF MATERIAL The material will be received in Type 6M shipping containers with less than 235 300 grams U per container. The pellets themselves will be sealed in plastic tubes with 20 to 30 pellets per tube. Wipes will be taken on the shipping container upon receipt, to test for contamination. The unopened drums will be locked in the SNM Storage Vault awaiting arrival of the customers' representative. The drums will be opened and wipes will be taken on the inner container and the outside of the plastic tubes. One or more of the tubes will be opened inside a chemistry hood equipped with a HEPA filtered exhaust. Wipes will be taken on the pellets to determine the amount of contamination to establish the appropriate handling procedures. The pellets will be put back in their tubes and returned to storage. 2.2 RADIATION SAFETY As a standard procedure, the safety rules contained in Appendix 1, Page 95, Section 1.1 " SAFETY" will be followed. April 20,1987 Page 117 l

For this program the radiation safety procedure listed below will be followed. 1. All personnel will wear film badges and dosimeters. 2. A contamination control area will be established and the work area will be i covered with protective paper. Catch trays and handling tools will be used to confine the pellets to the controlled area. 3. Rubber or plastic gloves and protective clothing will be worn as necessary. 4. Contamination surveys will be taken daily to ensure that there is no spread of contamination. 5. A portable air sampler will be operated in the work area during test rod loading and unloading operations. Air samples will be changed and analyzed each day. 2.3 TEST ROD LOADING AND UNLOADING PROCEDURES 1. Establish temporary controlled area with ropes and signs around the work l area. 2. Cover work area with protective paper. 3. Remove the plastic tubes containing the pellets for one rod from the storage area. 4. Open one plastic tube at a time and transfer the pellets to the test rod. Seal I empty plastic tube. 5. Continue loading the test rod with pellets until it is filled, then seal the end l cap onto the rod. 6. Return plastic tubes to shipping drum. { 7. Take wipes of work area and test rod. 8. Put test rod in storage area. 9. Set up to load the next rod. Repeat Steps 3 through 8 until all the test rods are loaded. 10. At completion of Rod Scanner Test Program, set up work area to unload the rods. 11. Remove plastic tubes from drum. April 20,1987 Page 118

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12. Break seal on test rod and return the pellets to the plastic tubes, sealing each tube as it is filled.

13. When all pellets are removed from the test rod and inserted in plastic tubes, take wipes of the plastic tubes, test rod, and work area.

14. Seal plastic tubes and place in the shipping container.

15. Repeat Steps 11 through 14 for all test rods.

2.4 SAFEGUARDS The quantity of material specified in this amendment is less than a " low strategic significant quantity" as defined in 10 CFR 73.2(y)(3); therefore, it will not be necessary to establish a Temporary Controlled Access Area with motion detector and/or surveil-lance. As a general rule of good safeguards practice, the fuel rods will be locked in the storage vault when not in use; however, they may be left on the rod scanner over night and during lunch hours for convenience. If the rods are not locked up, the test personnel who leave them out will be required to verify that the rods are still present when they return. 2.5 RADIOACTIVE MATERIAL SHIPPING PROCEDURES 1. Confirm that the recipient is authorized to receive the material. 2. Monitor and wipe the source or the radioactive material. 3. Determine the DOT quantity of the material. 4. Package in the appropriate inner container, if needed. l l S. Survey, wipe, and mark the inner container. 6. Package in the appropriate shipping container. 7. Survey, wipe, and apply the appropriate seal. 8. Apply the proper hazardous materials label (s) with the necessary information entered in the blanks on the label. ( 9. Fill out the Radioactive Materials Shipping Record. 'p 10. For SNM, fill out the proper DOE /NRC forms and send to the appropriate recipients. April 20,1987 Page 119 f !l. .. -.. - _..,. _. -... _.}}